The 13th International Conference on Environmental ... - Events

FINAL PROGRAM 

<strong>The</strong> <strong>13th</strong> <strong>International</strong> <strong>Conference</strong> on 

Environmental Remediation and 

Radioactive Waste Management 

October 3-7, 2010 – Tsukuba, Japan, Epochal Tsukuba 

http://www.jsme.or.jp/pes/ICEM10/ 

http://www.asmeconferences.org/ICEM2010/

ASME Message 

ICEM2010 is the thirteenth in a series of international conferences on 

environmental remediation and radioactive waste management organized by ASME, in 

cooperation with various agencies and other technical societies. While the sponsoring leads 

are the Environmental Engineering and the Nuclear Engineering Divisions within ASME, it 

provides a focus on environmental restoration and radioactive waste management for the 

entire ASME. Founded in 1880 as the American Society of Mechanical Engineers, today's 

ASME promotes the art, science & practice of mechanical & multidisciplinary engineering 

and allied sciences around the globe –with >120,000 members worldwide. 

Having chaired various ICEM conferences, I learned firsthand that success 

results only through the concerted effort of many individuals and organizations. If I tried 

to mention all, I would certainly fail, however, I must acknowledge the dedication of the 

<strong>Conference</strong> Secretary, Hiroyoshi Ueda, the continued support of the US Department of 

Energy, the Technical and Track Chairs, numerous volunteers and organizational sponsors, 

and most of all, the presenters and program participants. Thank you for being here. 

I first served as <strong>Conference</strong> Chair during ICEM ’01, which was held in Bruges, 

Belgium, during uncertain times for the world – just a few days after a terrorist attack of 

the United States. Since then the world has changed and the global environment has 

become much more personal. My personal message for ICEM’01 was of to collect the best 

of ideas, and “inclusion”. While not intentional, it became the first of a series of thematic, 

clear, and simple messages. 

ICEM returned to Bruges in 2007, and my message had two parts. I believe 

that they are even more valid today. First, that Education, Energy, Environment, and the 

Economy are inextricably intertwined, and therefore the logical development of one is 

strongly dependent on the health of the others. And second, that Society demands to be 

actively informed --that is, to be aware of the details that influence their reasonable 

expectation of an adequate return on their investment in Governance. In today's world, 

Global Environmental Partnerships and cooperative agreements are essential in order to 

demonstrate good Governance. 

<strong>The</strong> message for ICEM2010 is “A clean environment sustained by nuclear 

energy”. It is as simple as that. 

Please note that ICEM2011 will be held in Reims (France) on September 2011. 

A call for papers appears at: icemconf.com/2011CallforPapers/tabid/494/Default.aspx. I 

look forward to an intensive exchange of ideas, and to the exploration of new partnerships. 

After all that, the only thing left is to Welcome You to the ICEM2010 <strong>Conference</strong> on behalf 

of ASME. 

Aníbal L. Taboas 

ASME Fellow 

Technical Co-Chair, ICEM2010 

October 2010, Tsukuba, Japan

General Information 

Objectives and Background 

<strong>The</strong> <strong>13th</strong> <strong>International</strong> <strong>Conference</strong> on 

Environmental Remediation and Radioactive Waste 

Management (ICME2010) is the thirteenth in a 

series of international conferences organized by the 

ASME and other technical societies, and being held 

in Japan after an eleven year interval. ICEM was 

first held in Hong Kong in 1987, followed by Kyoto, 

Japan in 1989; Seoul, Korea, in 1991; Prague, Czech 

Republic, in 1993; Berlin, Germany, in 1995; 

Singapore in 1997; Nagoya, Japan in 1999; Bruges, 

Belgium in 2001 and 2007; Oxford, UK in 2003; 

Glasgow, Scotland in 2005; and Liverpool, UK in 

2009. 

ICEM promotes broad global exchange of 

information on technologies, operations, 

management approaches, economics and public 

policies in the critical areas of environmental 

remediation and radioactive waste management. 

ICEM also provides a unique opportunity to foster 

cooperation among specialists in mature 

environmental management programs and those 

involved in emerging programs. Attendees include 

scientists and engineers, suppliers and vendors, 

utilities, regulators government, and others seeking 

solution to environmental problems. Over 30 

countries are generally represented in ICEM 

conferences. 

<strong>Conference</strong> Co-Sponsors 

ICEM2010 is co-sponsored by: 

<strong>The</strong> Power and Energy System Division of the Japan 

Society of Mechanical Engineers (JSME), 

<strong>The</strong> Division of Nuclear Fuel Cycle and 

Environment of the Atomic Energy Society of Japan 

(AESJ), and 

<strong>The</strong> Environmental Engineering and the Nuclear 

Engineering Divisions of the American Society of 

Mechanical Engineers (ASME). 

Supporting Organizations 

<strong>The</strong> following national/international organizations 

support the conference: 

<strong>The</strong> Japan Atomic Energy Agency 

Tokyo Electric Power Company 

Nuclear Waste Management Organization of 

Japan 

Radioactive Waste Management Funding and 

Research Center 

1 

Tsukuba City 

<strong>The</strong> <strong>International</strong> Atomic Energy Agency 

<strong>The</strong> Canadian Nuclear Society 

<strong>The</strong> Chinese Nuclear Society 

<strong>The</strong> Korean Nuclear Society 

<strong>The</strong> Korean Radioactive Waste Society 

<strong>Conference</strong> Formant 

<strong>The</strong> ICEM2010 technical program includes 

concurrent sessions in six technical tracks: 

� Low/Intermediate-Level Radioactive Waste 

(L/ILW) 

� Spent Fuel, Fissile Material, Transuranic and 

High-Level Radioactive Waste (SF/TRU/HLW) 

� Facility Decontamination and 

Decommissioning (D&D) 

� Environmental Remediation (ER) 

� Environmental Management, Public 

Involvement and Crosscutting Issues (EM/PI) 

� Global Partnership and Multi-National 

Programs (GP) 

<strong>The</strong> technical program consists of an opening 

plenary session and several parallel program tracks 

with up to six concurrent sessions. <strong>The</strong> sessions 

include 25 minute oral presentations, panels, and 

poster displays which are designed to enhance 

dialogue between presenters and participants. <strong>The</strong> 

program is divided into 36 sessions conducted over 

three days. A listing of the specific sessions within 

each of the six technical program tracks can be 

found in the section for the technical sessions. 

<strong>The</strong> full registration fee includes meeting materials 

in the conference bag, entrance to all sessions and 

the exhibition, the welcome reception in the Sunday 

evening, lunch in the exhibition area and the banquet 

in the Tuesday evening. Each registrant will receive 

the conference proceedings on a CD-ROM mailed 

after the conference. 

<strong>Conference</strong> Venue 

ICEM2010 is being held at the Tsukuba 

<strong>International</strong> Congress Center, Epochal Tsukuba, 

which is home to many internationally renowned 

scientists. 

City of Tsukuba 

<strong>The</strong> City of Tsukuba is about 50 km northeast of 

Tokyo and 40 km northwest of Narita Airport. It 

covers an area of 28,000 ha with a population of 

about 200,000. To the north lies Mount Tsukuba,

877 meters above sea level. Seen from one angle, 

Tsukuba city, with its streets of traditional stores and 

houses, has the look of a rural refuge of abundant 

greenery. Seen from another angle, it is Tsukuba 

Science City, a superb modern urban landscape 

seldom seen in Japanese cities. 

<strong>Conference</strong> Check-In 

Upon your arrival, please check-in, or on-site 

register, at the Registration Desk and receive your 

own badge, i.e. registration pass, and other materials. 

It is strongly recommended that the conference 

participants check-in on Sunday, October 3, to avoid 

the rush before the opening session on Monday 

morning, October 4. Badges are required for the 

Welcome Reception in the Sunday evening. <strong>The</strong> 

Registration Desk is located in the entrance hall on 

the ground floor, during the following hours: 

Sunday, October 3 16:00 to 19:00 

Monday, October 4 7:30 to 17:00 

Tuesday, October 5 7:30 to 17:00 

Wednesday, October 6 7:30 to 13:00. 

Welcome Reception 

<strong>The</strong> Welcome Reception will be held from 18:00, 

Sunday, October 3, at the conference room 101+102 

of Epochal Tsukuba, the conference venue. 

Daily Morning Briefing for Speaker/Session 

Co-Chair 

All speakers including panelists and session 

co-chairs are invited to the Moring Briefing for 

Speaker/Session Co-Chair, on the day of their oral 

presentation/panel discussion/chairing. Coffee and 

bread will be served at the briefing. This briefing 

will give them the time for final arrangements 

before the oral session. <strong>The</strong> attendance at the 

briefing will provide an opportunity for the Session 

Co-Chairs to meet with the speakers, and for all to 

discuss the topics they will be addressing. It is 

essential that all speakers and session co-chairs 

attend the briefing. <strong>The</strong> briefing will be held at 

Rooms 401 through 404 during the following time 

slots: 



Wednesday, October 6 8:00 to 8:30. 

Lunch and Coffee in Exhibition Area 

Lunch will be served in the exhibition area, 

Multi-Purpose Hall on the ground floor. Lunches 

served on Monday, Tuesday and Wednesday will 

2 

mainly consist of a light meal that can be easily 

consumed while visiting the exhibits. <strong>The</strong> full 

registration fee includes the luncheon price for all 

three days. If you prefer, there are also several local 

restaurants near the conference venue. During the 

breaks, coffee/tea will be served in the exhibition 

area. 

Banquet 

<strong>The</strong> Banquet will be held from 18:30, Tuesday, 

October 5, at Banquet Hall "Subaru" on the ground 

floor of the annex building of Okura Frontier Hotel 

Tsukuba. 

Preparatory PC for Speakers 

<strong>The</strong> preparatory PC will be available in Room 301 

for speakers to load and review their PowerPoint 

slides. PC in each oral session room will be also 

available at the breaks for speakers to check their 

slides. 

Paid Business Services and Internet 

Paid Business Services are provided on the ground 

floor (See “Floor Map of Epochal Tsukuba and Paid 

Business Services” page.) Wireless Internet is 

available in the public area. 

Technical Tour 

All foreign participants are required to carry their 

passports to the tour for the security checks. A full 

day tour to Tokai-Mura, one of the well-known 

nuclear sites in Japan, will be held on October 7. 

<strong>The</strong> tour will integrate visits to ENTRY, JAEA’s 

research facility for the geological disposal, and the 

decommissioning site of a JAPC’s gas-cooled 

reactor, Tokai-1. (<strong>The</strong> tour is full.) 

Exhibition 

An industrial exhibition will be held in conjunction 

with the conference sessions during the following 

hours: 



Wednesday, October 6 9:00 to 13:30 

<strong>The</strong> Exhibition Area will be Multi-Purpose Hall on 

the ground floor.

Insurance and Liability 

All participants are encouraged to make their own 

arrangements for health and travel insurance. 

Neither JSME, AESJ, ASME nor their agents can be 

held responsible for any personal injury, loss, 

damage, accident to private property or additional 

expenses incurred because of delays or changes in 

air, rail, sea, road or other services, strikes, sickness, 

weather or any other cause. 

Disclaimer 

Neither JSME, AESJ nor ASME can accept any 

liability for death, injury, or any loss, cost or 

expense suffered or incurred by any person if such 

loss is caused or results from the act, default or 

omission of any person other than an employee or 

agent of JSME, AESJ or ASME. In particular, 

neither JSME, AESJ nor ASME can accept any 

liability for losses arising from the provision or 

non-provision of services provided by hotel 

companies or transport operators. Nor can JSME, 

AESJ nor ASME accept liability for losses suffered 

by reason of war including threat of war, riot and 

civil strife, terrorist activity, natural disaster, weather, 

flood, drought, technical, mechanical or electrical 

breakdown within any premises visited by delegates 

and/or partners in connection with the conference, 

industrial disputes, governmental action, regulations 

or technical problems which may affect the services 

provided in connection with the conference. Neither 

JSME, AESJ nor ASME is able to give any warranty 

that a particular person will appear as a speaker or 

panelist. 

3 

Technical Session Schedule 

Opening Session 

<strong>The</strong> opening session will start at 9:30 AM, Monday. 

Opening/welcome addresses from the conference 

organizers and plenary speeches from 

national/international representatives will be held. 

Each speech will be allocated 35 minutes including 

questions and answers. 

Concurrent Oral Sessions 

Concurrent sessions will be held in the six technical 

tracks. <strong>The</strong> sessions will start at 13:30 PM for 

Monday and at 9:00 AM for Tuesday and 

Wednesday. Each presentation will be allocated 25 

minutes (20 min. presentation and 5 min. discussion). 

<strong>The</strong> session co-chairs can extend the discussion time, 

keeping the time slot of the session. 

Poster Sessions 

<strong>The</strong> poster sessions will be held at Room 102 during 

the following time slots: 


17:25 to 17:50 


15:15 to 15:40 

<strong>The</strong> posters should be set up in Monday AM and 

torn down by 12:00, Wednesday.

List of Organization Acronyms 

AECL Atomic Energy of Canada Limited 

AESJ Atomic Energy Society of Japan 

AIST National Institute for Advanced Industrial Science and Technology 

ANL Argonne National Laboratory 

ANSTO Australian Nuclear Science and Technology Organisation 

ASME American Society of Mechanical Engineers 

CEA Commissariat à l'énergie atomique 

CRIEPI Central Research Institute of Electric Power Industry 

EDF Electricité de France 

EPRI Electric Power Research Institute 

IAE Institute of Applied Energy 

IAEA <strong>International</strong> Atomic Energy Agency 

JAEA Japan Atomic Energy Agency 

JAEA/NPSTC JAEA/Nuclear Nonproliferation Science and Technology Center 

JAPC Japan Atomic Power Company 

JNES Japan Nuclear Energy Safety Organization 

JNFL Japan Nuclear Fuel Limited 

JSME Japan Society of Mechanical Engineers 

KAERI Korea Atomic Energy Research Institute 

KAIST Korea Advanced Institute of Science and Technology 

KHNP Korea Hydro and Nuclear Power Company 

KNF Korea Nuclear Fuel 

KRMC Korea Radiactive Waste Management Corporation 

LANL Los Alamos National Laboratory 

LBNL Lawrence Berkeley National Laboratory 

MEXT Ministry of Education, Culture, Sports, Science and Technology 

NEL Nuclear Engineering, Ltd. 

NUMO Nuclear Waste Management Organization of Japan 

NWMO Nuclear Waste Management Organization 

OECD/NEA Organization for Economic Cooperation and Development/Nuclear Energy Agency 

PNNL Pacific Northwest National Laboratory 

RWMC Radioactive Waste Management Funding and Research Center 

SCK•CEN Studiecentrum voor Kernenergie•Centre d'Etude de l'Energie Nucléaire (Belgian 

Nuclear Research Centre) 

Tokyo Tech Tokyo Institute of Technology 

UCB University of California, Berkeley 

US DOE U. S. Department of Energy 

US NRC U. S. Nuclear Regulatory Commission 

VUB Vrije Universiteit Brussel 

4

Map around <strong>Conference</strong> Venue, Hotels and Tsukuba Station 

Okura Frontier Hotel 

Tsukuba Epochal 

Annex 

Okura Frontier Hotel Tsukuba 

(Epochal Tsukuba) 

5 

Main

Floor Map of Epochal Tsukuba and Paid Business Services 

6 

Copy, Printer, FAX 

Locker 

Mailing 

Beverage Vending Machine 

Information

ICEM2010 Technical Sessions at a Glance 

9:30 

9:45 

9:45 - 12:05 

12:10 - 13:30 

13:30 

13:35 

14:00 

14:25 

14:50 

15:15 - 15:40 

15:45 

16:10 

16:35 

17:00 

Monday, October 4 

Room 405 Room 101 Room 406 Room 202A Room 202B Room 303 

Preparation Preparation Preparation Preparation Preparation Preparation 

L1: Waste 

Management 

Break/Poster 

Break 

Preparation 

Main Convention Hall 

Opening Session 

Opening and Welcome Addresses 

Plenary Speeches from Representatives 

Break/Poster 

Preparation Preparation Preparation Preparation Preparation 

L1: Waste 

Management 

H1: National 

and 

<strong>International</strong> 

Programs (1) 

H3: Panel 

"Radwaste 

Human 

Resource 

Development 

to Support the 

Nuclear 

Renaissance" 

D1: National and 


D&D Programs 

D1: National and 


D&D Programs 

17:25 - 17:50 Poster Poster Poster Poster Poster Poster 

7 

Lunch 

R1: 

Environmental 

Impacts 

Break/Poster 

R1: 


Impacts 

M1: 


Management 

Break/Poster Break/Poster 

M1: 


Management 

H2: 

Transportation, 

Storage and 

Waste Treatment 

H2: 

Transportation, 

Storage and 

Waste Treatment


9:00 

9:05 

9:30 

9:55 

Room 201 Room 101 Room 202 Room 406 Room 405 Room 303 


10:20 - 10:45 Break/Poster Break/Poster Break/Poster Break/Poster Break/Poster Break/Poster 

10:50 

11:15 

11:40 

12:05 


12:30 - 13:30 Lunch Lunch Lunch Lunch Lunch Lunch 

13:35 

14:00 

14:25 

14:50 


15:15 - 15:40 Break/Poster Break/Poster Break/Poster Break/Poster Break/Poster Break/Poster 

15:45 

16:10 

16:35 

17:00 

17:25 

Tuesday, October 5 

L2: Solidification 

and Package 

(1) 


and Package (1) 


and Package (2) 

H4: National and 


Programs (2) 

H4: National and 


Programs (2) 

H5: Panel 

"Advances in 

Knowledge 

Management 

for Radioactive 

Waste 

Disposal" 

G2: IAEA 

Topical for 

Disused 

Sealed 

Radioactive 

Sources 

(DSRS) 

L3: Nuclide 

Assay 


L5: Recycling 

and Clearance 

H5: Panel 

"Advances in 

Knowledge 

Management 

for Radioactive 

Waste 

Disposal" 

D2: Dismantling 

and 

Decontamination 

D2: Dismantling 

and 

Decontamination 

D4: Panel 

"Applying 

Lessons 

Learned from 

Past D&D 

Activities" 

D4: Panel 

"Applying 

Lessons 

Learned from 

Past D&D 

Activities" 

8 

G1: 


Collaboration D3: 

Planning 

G2: IAEA 

Topical for 

Disused 

Sealed 

Radioactive 

Sources 

(DSRS) 

D3: 

Planning 

H6: Coupled 

Process 

Modeling and 

Natural 

Analogues 

H6: Coupled 

Process 

Modeling and 

Natural 

Analogues 

L3: Nuclide 

Assay 

R2: 


Remediation 

R2: 


Remediation


9:00 

9:05 

9:30 

9:55 

Room 201 Room 101 Room 406 Room 405 Room 303 


10:20 - 10:35 Break Break Break Break Break 

10:40 

11:05 

11:30 

11:55 


12:20 - 13:30 Lunch Lunch Lunch Lunch Lunch 

13:35 

14:00 

14:25 

14:50 


15:15 - 15:30 Break Break Break 

15:35 

16:00 

16:25 

16:50 

17:15 

Wednesday, October 6 

L6: Waste 

Treatment 

L6: Waste 

Treatment 

L7: Storage and 

Disposal Facility 

H7: Performance 

Assessment 

Modeling and 

Parameters 

H7: Performance 

Assessment 

Modeling and 

Parameters 

H9: Repository 

Engineering and 

Demonstration 

D5: 

Measurement 

and Estimation 

D6: Waste 

Treatment and 

Non-Reactor 

Preparation Preparation Preparation 

H9: Repository 

Engineering 

and 


D5: 

Measurement 

and Estimation 

D6: Waste 

Treatment and 

Non-Reactor 

9 

H8: Site 

Characterization 

and Modeling of 

Geological 

Environment (1) 

H8: Site 



Geological 


H10: Site 



Geological 


H10: Site 



Geological 


R3: ER 

Techniques 

R3: ER 

Techniques 

M2: Public 

Involvement

Monday, October 4, 2010 

OPENING SESSION 

Monday 09:30 Main Convention Hall 

--------------------------------------------------------------------------------------- 

Opening and Welcome Addresses: 

Satoru Tanaka, <strong>Conference</strong> General Chair 

Masanori Aritomi, Technical Program Chair 

Anibal L. Taboas, Technical Program Co-Chair, ASME 

Plenary Speeches: 

Tatsujiro Suzuki, Vice Chairman, Japan Atomic Energy 

Commission (Japan) 

“Nuclear Energy Strategy for Sustainable Growth: 

Aiming at Green Innovation and Life Innovation” 

Dae Y. Chung, Principal Deputy Assistant Secretary for 

Environmental Management, US DOE (USA) 

“U.S. Office of Environmental Management - World 

Leaders in Nuclear Cleanup and Construction” 

Ho Taek Yoon, Senior Vice President, KRMC (Korea Rep.) 

“Recent Progress in Radioactive Waste Management in 

Korea” 

Irena Mele, Head of Waste Technology Section, Division of 

Nuclear Fuel Cycle and Waste Technology, IAEA 

“Radioactive waste management – achievements, needs 

and future expectations” 

SESSION L1: Waste Management 

Monday 13:30 Room 405 

--------------------------------------------------------------------------------------- 

Session Co-Chairs: Miklos Garamszeghy, NWMO (Canada) 

and TBD 

1. 40081 – Radioactive Waste: Feedback of 40-year 

Operations in France 

Michel Dutzer, Gérald Ouzounian, Roberto Miguez, 

Jean-Louis Tison, Andra (France) 

2. 40227 – ILW Characterisation and Strategy Development 

at Chapelcross Power Station, Dumfriesshire, Scotland 

Rob Hodgson, Alastair Stewart, Greg Wotherspoon,� 

Magnox North � (UK); Vince Cane, Dr Rob Thied, 

Nuclear Technologies PLC� (UK) 

3. 40226 – <strong>The</strong> Ethics of the Management of Low and 

Intermediate Radioactive Wastes Generated by Cernavoda 

NPP, a Challenge for the Romanian Specialists. 

Gheorghe Barariu, Subsidiary of Technology and 

Engineering for Nuclear Objectives (Romania) 

4. 40149 – CEA's radioactive waste and unused fuel 

inventory - Marcoule site example 

Jean-Guy Nokhamzon, Marc Butez, Ddaniel Fulleringer, 

CEA (France) 

——————— Break ——————— 

5. 40258 – Management of historical waste legacy at NRG 

Petten 

Renate de Vos, Nuclear Research and Consultancy Group� 

(Netherlands) 

6. 40031 – Norwegian Support For Regulations Of 

Radioactive Waste Management From Uranium Mining 

And Mill Tailings In Central Asia 

Tamara Zhunussova, Malgorzata Sneve, Astrid Liland, 

Norwegian Radiation Protection Authority (Norway); 

Alexander Kim, Kazakhstan Atomic Energy Committee 

Technical Sessions 

10 

(Kazakhstan); Ulmas Mirsaidov, Tajikistan Nuclear and 

Radiation Safety Agency (Tajikistan);Baigabyl 

Tolongutov, Chui Ecological Laboratory of Kyrgiz 

Republic (Kyrgiz); Per Strand, Norwegian Radiation 

Protection Authority, University of Life Sciences (UMB) 

(Norway) 

7. 40259 – Strengthening the Nuclear Security System of 

Central Radioactive Waste Processing and Storage 

Facility at AERE 

Abdus Sattar Mollah, Bangladesh Atomic Energy 

Commission (Bangladesh) 

SESSION H1: : National and <strong>International</strong> Programs (1) 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Stratis Vomvoris, Nagra (Switzerland) 

and Hiromi Tanabe, RWMC (Japan) 

1. 40097 – Overview of NUMO's policy for implementing 

safe geological disposal and developing supporting 

technologies 

Hiroyuki Tsuchi, Kenichi Kaku, Katsuhiko Ishiguro, Akira 

Deguchi, Yoshiaki Takahashi, NUMO (Japan) 

2. 40150 – Stepwise Site Selection in Switzerland - Sectoral 

Plan: Status and Outlook 

Thomas Ernst, Stratis Vomvoris, Markus Fritschi, Nagra 

(Switzerland) 

3. 40084– Site Selection for a Geological Disposal in 

France: an Approach of Convergence 

Gérald� Ouzounian, Roberto Miguez, Jean-Louis Tison, 

Andra (France) 

SESSION H2: Transportation, Storage and Waste 

Treatment 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Hiroshige Kikura, Tokyo Tech (Japan) 

and Anibal L. Taboas, ANL (USA) 

1. 40155 – Support of the Nuclear Research Institute Rez plc 

of the Shipment of Spent Nuclear Fuel from Research 

Reactors to the Russian Federation for Reprocessing in 

the Frame of the RRRFR Program 

Josef Podlaha, Karel Svoboda, Nuclear research 

institute plc. (Czech Republic) 

2. 40177 – Long Term Storage of Nuclear Spent Fuel as Key 

Role of Japan's Nuclear Fuel Cycle until 2100: Cost and 

Benefit 

Tadahiro Katsuta, Meiji University (Japan) 

3. 40285 – TRU Recycling Options for increasing Protected 

Plutonium Production of FBR 

Sidik Permana, Mitsutoshi Suzuki, JAEA (Japan) 

4. 40132 – Transuranic (TRU) Waste Volume Reduction 

Operation at a Plutonium Facility 

Michael Cournoyer, Archie E. Nixon, Keith W. Fife, 

Arnold M. Sandoval, Vincent E. Garcia, Robert L. Dodge, 

LANL (USA) 

——————— Break ——————— 

5. 40247 – Disposition of Transuranic Residues from 

Plutonium Isentropic Compression Experiment (Pu-ICE). 

LA-UR 10-04699. 

Kapil Goyal, David M. French, LANL (USA); Betty J. 

Humphrey, Weston Solutions, Inc. (USA); Jeffry Gluth, 

Sandia National Laboratories/NM (USA) 

6. 40188 – A Milestone in Vitrification - the Replacement of 

a Hot Metallic Crucible with a Cold Crucible Melter in a 

Hot Cell at the La Hague Plant

Sphiee Robert, Benoit Carpentier, SGN (France); 

Florence Gassot Guilbert, SGN Service procédé 

(France); Sandrine Naline, AREVA (France); Frédéric 

Gouyaud, AREVA NC (France); Christophe Girold, CEA 

(France) 

7. 40265 – Adaptation of CCIM technology for HLW 

treatment. Results of research and development. 

Vladimir Lebedev, Alexander Kobelev, Fyodor Lifanov, 

Sergey Dmitriev, SIA Radon (Russia) 

SESSION H3: Panel "Radwaste Human Resource 

Development to Support the Nuclear Renaissance" 


--------------------------------------------------------------------------------------- 

Session Chair: Ian G. McKinley, McKinley Consulting 

(Switzerland) 

<strong>The</strong> panel focuses on answering three questions that form the 

basis for establishing an efficient and effective human resource 

development plan for radioactive waste management: 

- what resources do we need and when do we need them? 

- how can we assure that these resources will be available 

(with emphasis on training requirements, especially for 

generalists and multidisciplinary coordinators)? 

- what infrastructure and procedures need to be made 

available with high priority now? 

Panelists: 

Joonhong Ahn, UCB (USA) 

Tomio Kawata, NUMO (Japan) 

Ian G. McKinley, McKinley Consulting (Switzerland) 

Irena Mele, IAEA 

Shawn Smith, US NRC (USA) 

SESSION D1 : National and <strong>International</strong> D&D 

Programs 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Sean Bushart, EPRI (USA) and Satoshi 

Yanagihara, JAEA (Japan) 

1. 40003 – Westinghouse PWR and BWR Reactor Vessel 

Segmentation Experience in Using Mechanical Cutting 

Process 

Joseph Boucau, Stefan Fallström, Per Segerud, Paul 

Kreitman, Westinghouse Electric Belgium (Belgium) 

2. 40130 – EPRI Nuclear Power Plant Decommissioning 

Technology Program 

Karen Kim, Sean Bushart, Mike Naughton, Richard 

McGrath, Electrict Power Research Institute (USA) 

3. 40253 – Tokai-1 Decommissioning Project 

Keizaburou Yoshino, JAPC (Japan) 

4. 40289 – Activities of the OECD/NEA WPDD in the Field 

of Decommissioning� 

Claudio Pescatore, Patrick O'Sullivan, OECD/NEA� 

—————— Break ——————— 

5. 40307 – French Decommissioning Feedback Experience 

and Lessons Learned 

Jean-Guy Nokhamzon, CEA (France); Patrick 

O'Sullivan, OECD/NEA 

6. 40032 – Decommissioning Planning for Swedish 

Operating NPPs � Gunnar Hedin, Mathias Edelborg, 

Niklas Bergh, Westinghouse Electric Sweden (Sweden); 

Jan Carlsson, Fredrik de la Gardie, SKB (Sweden) 

7. 40104 – Chernobyl nuclear power plant 

Decommissioning Program 

Denys Tkachov, Dmytro Stelmakh, Viktor Kuchinskyy, 


11 

State Special Enterprise Chernobyl Nuclear Power Plant 

(Ukraine) 

8. 40273 – Considerations for Grout Formulations in 

Facility Closures using In Situ Strategies� 

John Gladen, Mike Serrato, Chris Langton, Savannah 

River National Laboratory (USA); Andy Szilagyi, US 

DOE (USA) 

SESSION R1: Environmental Impacts 

Monday 13:30 Room 202A 

--------------------------------------------------------------------------------------- 

Session Co-Chairs: Shinzo Ueta, Mitsubishi Materials 

Corporation (Japan) and Dawn Wellman, PNNL (USA) 

1. 40298 – Main Results of A Remediation of Uranium- and 

CHC-Contaminated Groundwater� 

Jörg Wörner, Sonja Margraf, Walter Hackel,� 

RD-Hanau (Germany) 

2. 40267 – Biogeochemical Gradients, Waste Site Evolution, 

and Implications for Sustained Metal and Radionuclide 

Attenuation in Complex Subsurface Environments� 

Karen Skubal, Justin Marble, Kurt D. Gerdes, US DOE 

(USA); Miles Denham, Karen Vangelas, Savannah River 

National Laboratory (USA) 

3. 40260 – Current Mercury Distribution and Bioavailability 

in Floodplain Soils of Lower East Fork Popular Creek, 

Oak Ridge, Tennessee, USA� 

Fengxiang X. Han, Yi Su, David L. Monts, Mississippi 

State University (USA) 

4. 40262 – Integrated Strategy to Address Hanford's Deep 

Vadose Zone Remediation Challenges� 

Mark B. Triplett, Mark D. Freshley, Michael J. Truex, 

Dawn Wellman, PNNL (USA); Kurt D. Gerdes, Briant L. 

Charboneau, John G. Morse, Robert W. Lober, US DOE 

(USA); Glen B. Chronister, CH2M Hill Plateau 

Remediation Company (USA) 

—————— Break ——————— 

5. 40235 – Advanced Remedial Methods for Metals and 

Radionuclides in Deep Vadose Zone Environments 

Dawn Wellman, Shas Mattigod, Ann Miracle, Lirong 

Zhong, Danielle Jansik, PNNL (USA); Susan Hubbard, 

Yuxin Wu, LBNL (USA); Martin Foote, MSE (USA) 

6. 40152 – Water discharge from the Chernobyl NPP cooling 

pond: the main challenges of the environmental protection 

during post-rehabilitation period 

Dmitri Gudkov, Institute of Hydrobiology (Ukraine) 

SESSION M1: Environmental Management 

Monday 13:30 Room 202B 

--------------------------------------------------------------------------------------- 

Session Co-Chairs: Motoi Kawanishi, CRIEPI (Japan) and 

Tadao Tanaka, JAEA (Japan) 

1. 40086 – Legacy Management: Turning Liabilities into 

Assets 

Joe Legare, Eric Olson, S.M. Stoller Corporation (USA) 

2. 40270 – Lessons Learned in Planning the Canadian 

Nuclear Legacy Liabilities Program 

Michael Stephens, Sheila M. Brooks, Joan Miller, Robert 

Mason, AECL (Canada) 

3. 40292 – GIS-Based Decision Support System for the 

Former Semipalatinsk Test Site 

Maira Mukusheva, Sergey Baranov, National Nuclear 

Center (Kazakhstan) 

4. 40218 – RFID Technology for Environmental 

Remediation and Radioactive Waste Management 

Hanchung Tsai, Yung Liu, ANL (USA); James Shuler, US

DOE (USA) 

—————— Break ——————— 

5. 40181 – <strong>The</strong> Radioactivity of 3H in Metals by a High 

Temperature Furnace and a Liquid Scintillation Counter� 

Hee Reyoung Kim, Geun Sik Choi, Sang Yun Park, 

Chang Woo Lee, Moon Hee Han, KAERI (Korea Rep.) 

6. 40275 – Next Generation Waste Glass Melters in the U.S. 

DOE Waste Processing Program� 

Steven P. Schneider, Gary Smith, US DOE (USA) 

Tuesday, October 5, 2010 

SESSION L2: Solidification and Package (1) 

Tuesday 09:00 Room 201 

--------------------------------------------------------------------------------------- 

Session Co-Chairs: Miklos Garamszeghy, NWMO (Canada) 

and Yoshihiko Horikawa, NEL (Japan) 

1. 40021 – Commercialization Project of Ulchin 

2. 

Vitrification 

Hyun-jun Jo, Cheon-Woo Kim, KHNP (Korea Rep.); 

Tae-Won Hwang, Nuclear Engineering & Technology 

Institure (Korea Rep.) 

40023 – Plasma Gasification/Vitrification of Wet ILW� 

Gary Hanus, John Williams, Matt Zirbes, Phoenix 

Solutions Co. (USA) 

3. 40026 – Solidification of Simulated Liquid Waste of 

Primary Loop Resin Elution Process of PWR 

Masamichi Obata, Michitaka Saso, Masaaki Kaneko, 

Nobuhito Ogaki, Taichi Horimoto, Toshiba corporation, 

Toshikazu Waki, <strong>The</strong> Kansai Electirc Power Co., Inc. 

(Japan) 

—————— Break ——————— 

4. 40108 – VUJE Experience with Cementation of Liquid 

and Wet Radioactive Waste� 

Kamil Kravárik, .Zuzana Holická, Anton Pekár, Milan 

Žatkulák, VUJE, Inc. (Slovakia) 

5. 40112 – Study of LPOP residue on spent-resin 

mineralization and solidification� 

Gen-ichi Katagiri, Morio Fujisawa, Fuji Electric Systems 

Co., Ltd. (Japan); Kazuya Sano, Norikazu Higashiura, 

JAEA (Japan) 

SESSION L3: Nuclide Assay 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: David James, DW James Consulting 

(USA) and TBD 

1. 40279 – Design of a facility for the automated receipt 

inspection and characterization of LILW using integrated 

non-destructive examination and assay techniques. 

Stephen Halliwell, VJ Technologies Inc. (USA) 

2. 40167 – Feasibility Study on the Nuclide Analysis of the 

Radwaste Drum Using the Spectrum to Dose Conversion 

Factor� 

Young-Yong Ji, Dae-Seok Hong, Tae-kuk Kim, Woo-Seog 

Ryu, KAERI (Korea Rep.) 

3. 40255 – Portable Non-Destructive Assay Methods for 

Screening and Segregation of Radioactive Waste 

Alan Simpson, Martin Clapham, Stephanie Jones, Randy 

Lucero, Pajarito Scientific Corporation (UK) 

—————— Break ——————— 

4. 40093 – Alpha radioactivity monitor using ionized air 

transportation for large size uranium waste (1) - Large 

measurement chamber and evaluation of detection 


12 

performance – 

Susumu Naito, Shuji Yamamoto, Miikio Izumi, Yosuke 

Hirata, Yukio Yoshimura, Tatsuyuki Maekawa, Toshiba 

Corporation (Japan) 

5. 40091 – Alpha Radioactivity Monitor using Ionized Air 

Transport Technology for Large Size Uranium Waste (2) - 

Simulation model reinforcement for practical apparatus 

design – 

Takatoshi Asada, Yosuke Hirata, Susumu Naito, Miikio 

Izumi, Yukio Yoshimura, Toshiba Corpoation (Japan) 

6. 40111 – Preparation of Reference Materials on 

Radiochemical Analysis for Low-Level Radioactive 

Waste Generated from Japan Atomic Energy Agency 

Ken-ichiro Ishimori, Yutaka Kameo, Mikio Nakashima, 

Kuniaki Takahashi, JAEA (Japan) 

SESSION H4: National and <strong>International</strong> Programs (2) 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Joonhong Ahn, UCB (USA) and Kenichi 

Kaku NUMO (Japan) 

1. 40213 – U.S. NRC Integrated Spent Fuel Management 

Plan 

Catherine Haney, Shawn Smith, US NRC (USA) 

2. 40116 – Regulatory Research for Geological Disposal of 

High-level Radioactive Waste in Japan� 

Shinichi Nakayama, JAEA (Japan); Yoshio Watanabe, 

AIST (Japan); Masami Kato, JNES (Japan) 

3. 40280 – Recent Developments and Trends on 

Requirements Management Systems� 

Satoru Suzuki, Hiroyoshi Ueda, Kiyoshi Fujisaki, 

Katsuhiko Ishiguro, Hiroyuki Tsuchi, NUMO (Japan); 

Stratis Vomvoris, Irina Gaus, Nagra (Switzerland) 

——————— Break ——————— 

4. 40228 – Development of Requirements Management 

System of NUMO and practical experience with 

development of the database contents 

Satoru Suzuki, Hiroyoshi Ueda, Katsuhiko Ishiguro, 

Hiroyuki Tsuchi, Kiyoshi Fujisaki, NUMO (Japan); 

Kiyoshi Oyamada, JGC Corporation (Japan); Shoko 

Yashio, Obayashi Corporation (Japan) 

5. 40231 – Application Of Lifecycle Management To Design 

of <strong>The</strong> UK Geological Disposal Facility 

Henry O'Grady, Malcolm Currie, Parsons 

Brinckerhoff Ltd. (UK); Philip Rendell, Radioactive, 

Waste Management Directorate (UK) 

SESSION D2: Dismantling and Decontamination 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Jean-Guy Nokhamzon, CEA (France) 

and Motonori Nakagami, Chubu Electric Power (Japan) 

1. 40036 – AREVA NP: Experience in Dismantling and 

Packaging of Pressure Vessel and Core Internals 

Peter Pillokat, Jan Hendrik Bruhn, AREVA NP GmbH 

(Germany) 

2. 40102 – Study on evaluation models of management data 

for decommissioning of Fugen 

Yuji Shibahara, Masanori Izumi, Takashi Nanko, Mitsuo 

Tachibana, Tsutomu Ishigami, JAEA (Japan) 

3. 40083 – CORD Decontamination Technologies for 

Decommissioning - A Comprehensive Approach Based on 

Over 30 Years Experience 

Christoph Stiepani, AREVA NP GmbH (Germany) 

——————— Break ———————

4. 40007 – Chemical Decontamination for 

Decommissioning (DFD) and DFDX� 

Ronald Morris, Westinghouse Electric Company (USA) 

5. 40127 – Methods for Calculation and Optimisation of 

Personnel Exposure during Planning of Decommissioning 

of Nuclear Installation� 

Marek Vasko, Ivan Rehak, DECOM, a.s. (Slovakia); 

Vladimir Daniska, DECONTA, a.s. (Slovakia); Vladimir 

Necas, Slovak University of Technology in Bratislava 

(Slovakia) 

SESSION D3: Planning 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Joseph Boucau, Westinghouse Electric 

Belgium (Belgium) and Toshihiko Higashi, Kansai Electric 

Power (Japan) 

1. 40129 – Program Change Management During Nuclear 

Power Plant Decommissioning 

Mike Naughton, Sean Bushart, Karen Kim, EPRI (USA) 

2. 40245 – Status of the Support Researches for the 

Regulation of Nuclear Facilities Decommissioning in 

Japan 

Yusuke Masuda, Yukihiro Iguchi, Satoru Kawasaki, 

Masami Kato, JNES (Japan) 

3. 40136 – Decommissioning Costing Approach Based on 

the Standardised List of Costing Items; Lessons Learnt 

Vladimir Daniska, Frantisek Ondra, Peter Bezak, 

DECONTA, a.s. (Slovakia); Ivan Rehak, Marek Vasko, 

Jozef Pritrsky, DECOM a.s. (Slovakia) ; Matej Zachar, 

Vladimir Necas, Slovak University of Technology in 

Bratislava (Slovakia) 

—————— Break ——————— 

4. 40290 – <strong>The</strong> Outline of Decommissioning Plan for 

Hamaoka Nuclear Power Station Unit-1 and Unit-2 

Yoshifusa Fukuoka, Chubu Electric Power Co., Inc. 

(Japan) 

5. 40015 – Study on Influence of Nuclear Fuel Material 

Management and Transfer Scenarios on 

Decommissioning 

Kazuma Mizukoshi, Nuclear Engineering, Ltd. (Japan) 

6. 40100 – Dose Assessment for setting of EPZ in 

Emergency Plan for Decommissioning of Nuclear Power 

Plant 

Hirokazu Minato, Hitachi-GE Nuclear Energy (Japan); 

Takatoshi Hattori, CRIEPI (Japan); Toshihiko Higashi, 

<strong>The</strong> Kansai Electric Power Co., Inc. (Japan); Takehiro 

Iwata, JAPC (Japan) 

7. 40044 – Decommissioning of Ignalina NPP 

Karolis Zemkajus, Algirdas Vaidotas, Radioactive Waste 

Management Agency in Lithuania (Lithuania) 

SESSION G1 : <strong>International</strong> Collaboration 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Hiromi Tanabe, RWMC (Japan) and 

Robin Heard, IAEA 

1. 40118 – Advancing the Use of IAEA Networks in 

Radioactive Waste Management: Past Successes, Present 

Challenges and Future Opportunities. 

Paul Degnan, John Kinker, Irena Mele, Paul J. Dinner, 

Horst Monken Fernandes, Antonio Morales, Lumir 

Nachmilner, Shaheed Hossain, IAEA 

2. 40287 – <strong>The</strong> activities of the OECD/NEA RWMC in the 

Field of HLW and SF disposal 


13 

Claudio Pescatore, OECD/NEA 

3. 40147 – Grimsel Test Site - Phase VI Status and Outlook 

Ingo Blechschmidt, Sven Peter Teodori, Stratis Vomvoris, 

Nagra (Switzerland) 



--------------------------------------------------------------------------------------- 

Session Co-Chairs: Gérald Ouzounian, Andra (France) and 

TBD 

1. 40299 – Treatment of low level radioactive waste by 

plasma: a proven technology? 

Jan Deckers, Belgoprocess, NV (Belgium) 

2. 40128 – <strong>The</strong> Zwilag Plasma Facility - Five Years of 

Successful Operation 

Walter Heep, Zwilag Interim Storage Facility 

(Switzerland) 

3. 40293 – Safety Assessment Of Disposal Container For 

Higher Activity Low Level Waste 

Motonori Nakagami, Seiji Komatsuki, Chubu Electric 

Power Co., Inc. (Japan); Kyosuke Fujisawa, Takashi 

Nishio, KOBE STEEL, LTD. (Japan); Thomas Quercetti, 

André Musolff, Karsten Müller, Federal Institute for 

Materials Research and Testing (Germany) 

SESSION L5: Recycling and Clearance 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Kapila Fernando, ANSTO (Australia) 

and TBD 

1. 40223 – NPP Bulk Equipment Dismantling Problems and 

Experience 

Alexander B. Gelbutovsky, Eugeny V. Balushkin, Jury A. 

Epikhin, Alexander V. Troshev, Peter I. Cheremisin, 

ECOMET-S JSC (Russia) 

2. 40073 – Reuse Of Conditionally Released Radioactive 

Materials From NPP Decommissioning Applied In 

Motorway Bridges Construction 

Michal Panik, Tomas Hrncir, Vladimir Necas, Slovak 

University of Technology in Bratislava (Slovakia) 

3. 40071 – Modelling of Motorway Tunnels Scenario for 

Utilization of Conditionally Released Radioactive 

Materials 

Tomas Hrncir, Michal Panik, Vladimir Necas, Slovak 

University of Technology in Bratislava (Slovakia) 

4. 40117 – Estimate of Clearance Levels for Metal Materials 

Contaminated with Uranium 

Seiji Takeda, Hideo Kimura, JAEA (Japan) 

SESSION H5: Panel "Advances in Knowledge 

Management for Radioactive Waste Disposal" 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Hiroyuki Umeki, Hitoshi Makino, JAEA 

(Japan) and Ian G. McKinley, McKinley Consulting 

(Switzerland) 

Part 1: Status and plans of KM activities in national 

programmes 

Information exchange with emphasis on identification of 

potential areas for collaboration. Panelists represent 

implementing organisations, regulatory authorities and R&D 

organisations, thus bringing different perspectives on KM. 

Part 2: Brainstorming on advanced KM tools

A common requirement in all geological disposal programmes 

is efficiently and rigorously managing increasing large and 

complex fluxes of information. Emphasis will be on suggesting 

practical applications and further improvements of advanced 

KM tools that have been developed by JAEA, distinguishing 

between programme-specific constraints and more generic areas, 

which could be a focus for future collaborative projects. 

Panelists: 

Johan Andersson, JA Streamflow (Sweden) 

Kenzi Karasaki, LBNL (USA) 

Masami Kato, JNES (Japan) 

Lawrence Kokajko, US NRC (USA) 

Mark Nutt, ANL (USA) 

Richard Shaw, BGS (UK) 

Hiroyuki Tsuchi, NUMO (Japan) 

Hiroyuki Umeki, JAEA (Japan) 

SESSION H6: Coupled Process Modeling and Natural 

Analogues 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Irina Gaus, Nagra (Switzerland) and 

Gento Kamei, JAEA (Japan) 

1. 40306 Keynote – A Discussion of Key Issues in Coupled 

THM Processes in Clays, Rock Salt and Crystalline Rock 

with Bentonite Buffer 

Chin-Fu Tsang, LBNL (USA) 

2. 40159 – Environmental Remediation of High-Level 

Nuclear Waste in Geological Repository: Modified 

Computer Code Creates Ultimate Benchmark in Natural 

systems 

Geoffrey Peter, Oregon Institute of Technology Portland 

Center (USA) 

3. 40196 – Effect of the Residual Heat Release of the 

Nuclear Waste Stored in an Unsaturated Zone on 

Radionuclide Release 

Lubna K. Hamdan, John C. Walton, Arturo Woocay, 

University of Texas at El paso (USA) 

4. 40072 – Evaluation of behavior of rare earth elements 

based on determination of chemical state in groundwater 

in granite 

Yuhei Yamamoto, Daisuke Aosai, Takashi Mizuno, JAEA 

(Japan) 

—————— Break ——————— 

5. 40022 – Natural analogue studies of bentonite reaction 

under hyperalkaline conditions: overview of ongoing 

work at the Zambales Ophiolite, Philippines 

Naoki Fujii, M. Yamakawa, RWMC (Japan); K. Namiki, 

Obayashi Corporation (Japan); T. Sato, Hokkaido 

University (Japan); N. Shikazono, Keio University 

(Japan); C. A. Arcilla, C. Pascua, University of the 

Philippines (Philippines); W Russell Alexander, 

Bedrock Geosciences ( Switzerland) 

6. 40063 – Natural Analogues of Cement: Overview of the 

Unique Systems in Jordan 

Gento Kamei, JAEA (Japan); W Russell Alexander, 

Bedrock Geosciences (Switzerland); Ian D. Clark, 

University of Ottawa (Canada); Paul Degnan, IAEA; 

Marcel Elie, Shell (Netherlands); Hani Khoury, Elias 

Salameh, University of Jordan (Jordan), Antoni E. 

Milodowski, British Geological Survey (UK), Alister F. 

Pitty, Pitty Consulting � (UK); John A.T. Smellie, 

Conterra (Sweden) 


14 

SESSION D4: Panel "Applying Lessons Learned from 

Past D&D Activities" 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Koji Okamoto, University of Tokyo 

(Japan) and Claudio Pescatore, OECD/NEA 

In this panel, international lessons learned from past D&D 

activities will be discussed to optimize future decommissioning 

activities. Each panelist will talk about some of topics below. 

Discussion will be focused on two or three topics. Later in the 

panel, a recent TV report in Japan will be introduced to discuss 

mass media issue, too. 

a. Decommissioning Project Management 

b. Decommissioning Techniques 

c. Waste Disposal 

d. Building and Site Remediation 

e. Social Issues and Others 

Panelists: 

Koji Okamoto, Keynote, University of Tokyo (Japan) 

Sean Bushart, EPRI (USA) 

Satoshi Karigome, JAPC (Japan) 

Jean-Guy Nokhamzon, CEA (France) 


Andy Szilagyi, US DOE (USA) 

Satoshi Yanagihara, JAEA (Japan) 

SESSION R2: Environmental Remediation 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Hirofumi Tsukada, Institute for 

Environmental Sciences (Japan) and Jörg Wörner, 

RD-Hanau (Germany) 

1. 40261 – Reclamation of Three In Situ Uranium Mines - 

Innovative Techniques 

Wallace Mays, W M Mining Company (USA) 

2. 40005 – Environmental remediation Activities at the 

Ningyo-toge Uranium Mine, Japan 

Hiroshi Saito, Tomihiro Taki, JAEA (Japan) 

3. 40092 – Radon impact at a remediated uranium mine site 

in Japan 

Yuu Ishimori, JAEA (Japan) 

4. 40243 – Phosphate based remediation techniques: 

interaction of phosphate with uranium-bound calcite 

Chase Bovaird, Dawn Wellman, PNNL (USA) 

—————— Break ——————— 

5. 40220 – Remediation of Old Environmental Liabilities in 

the Nuclear Research Institute Rez plc 

Karel Svoboda, Josef Podlaha, Nuclear Research Institute 

Rez plc (Czech Republic) 

SESSION G2: IAEA Topical for Disused Sealed 

Radioactive Sources (DSRS) 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Hiromi Tanabe, RWMC (Japan) and 

Irena Mele, IAEA 

1. 40028 – <strong>International</strong> initiatives addressing the safety and 

security of Disused Sealed Radioactive Sources (DSRS) 


2. 40303 – Current situation and Management Plan of 

Radioactive Sources in Japan 

Hirokuni Ito, Tadashi Ishii, Tomokazu Ueta, Takao 

Nakaya, Kenya Suyama, MEXT (Japan)

3. 40060 – <strong>The</strong> Deployment of the Mobile Hot Cell to 

Condition High Activity Disused Sealed Radioactive 

Sources (DSRS) for Long Term Storage or Removal 

Gerhardus R. Liebenberg, South African Nuclear Energy 

Corporation (Necsa) (South Africa) 

4. 40266 – Problems with Packaged Sources in Foreign 

Countries 

James Matzke, John Zarling, Cristy Abeyta, Joseph A. 

Tompkins, LANL (USA) 

—————— Break ——————— 

5. 40085 – <strong>The</strong> IAEA's approach to the security of 

radioactive material 


6. 40058 – Radioactive Waste Management in Lebanon 

Munzna Assi, Lebanese Atomic Energy Commission 

(Lebanon) 

7. 40029 – <strong>The</strong> ultimate solution – disposal of Disused 

Sealed Radioactive Sources (DSRS) 


Wednesday, October 6, 2010 

SESSION L6: Waste Treatment 

Wednesday 09:00 Room 201 

--------------------------------------------------------------------------------------- 

Session Co-Chairs: Gérald Ouzounian, Andra (France) and 

Hirokazu Tanaka, Mitsubishi Materials Corporation 

(Japan) 

1. 40055 – Drying System For Radioactivated Metal Waste 

From Nuclear Power Station 

Nobuhito Ogaki, Yasushi Ooishi, Hironori Takabayashi, 

Masamichi Obata, Taichi Horimoto, Toshiba Corporation 

(Japan) 

2. 40186 – Macroporous Catalysts for Hydrothermal 

Oxidation of Metallorganic Complexes at Liquid 

Radioactive Waste Treatment 

Valentin Avramenko, Dmitry Marinin, Vitaly Mayorov, 

Alexander Mironenko, Marina Palamarchuk, Valentin 

Sergienko, Institute of Chemistry FEDRAS (Russia) 

3. 40157 – Carbonaceous radioactive waste treatment 

Gerard Laurent, EDF/CIDEN (France) 

—————— Break ——————— 

4. 40163 – Impermeable graphite: A new development for 

the waste management of irradiated graphite 

Johannes Fachinger, Karl-Heinz Grosse, Furnances 

Nuclear Applications Grenoble (Germany); Richard 

Seemann, Milan Hrovat ALD (Germany) 

5. 40165 – THOR® Steam Reforming Technology for the 

Treatment of Ion Exchange Resins and More Complex 

Wastes such as Fuel Reprocessing Wastes 

J. Brad Mason, Corey Myers, Studsvik, Inc. (USA) 

6. 40257 – Phase Behavior and Reverse Micelle Formation 

in Supercritical CO2 with DTAB and F-pentanol for 

Decontamination of Radioactive Wastes 

Kensuke Kurahashi, Osamu Tomioka, Yoshihiro Meguro, 

JAEA (Japan) 

SESSION H7: Performance Assessment Modeling and 

Parameters 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Hiroyuki Umeki, JAEA (Japan) and 

Masaki Tsukamoto, CRIEPI (Japan) 


15 

1. 40305 Keynote – Development of a Realistic Repository 

Performance Assessment Method 


2. 40204 – Integrated model for the near field of a repository 

in granite host-rock. Probabilistic approach 

Lara Duro, Alba Valls, Olga Riba, Jordi Bruno, Amphos 

XXI Consulting S.L.(Spain); Aurora Martinez-Esparza, 

ENRESA (Spain) 

3. 40017 – Spatial Variability and Parametric Uncertainty in 

Performance Assessment Models 

Osvaldo Pensado, James Mancillas, Scott Painter, 

Southwest Research Institute (USA); Yasuo Tomishima, 

AIST (Japan) 

—————— Break ——————— 

4. 40203 – Development of a Radiolytic Model for the 

Alteration of Spent Nuclear Fuel. Incorporation of 

non-oxidative matrix dissolution and hydrogen 

oxidation inhibition effect 

Lara Duro, Alba Valls, Olga Riba, Jordi Bruno, Amphos 

XXI Consulting S.L.(Spain); Aurora Martinez-Esparza, 

ENRESA (Spain) 

5. 40172 – Evaluated and Estimated Solubility of Some 

Elements for Performance Assessment of Geological 

Disposal of High-level Radioactive Waste Using Updated 

Version of <strong>The</strong>rmodynamic Database 

Akira Kitamura, Reisuke Doi, JAEA (Japan); Yasushi 

Yoshida, Inspection Development Co., Ltd. (Japan) 

6. 40049 – Consideration on Soil Origin Carbon Transfer to 

Leafy Vegetables Using Stable Carbon Isotope Ratios 

Keiko Tagami, Shigeo Uchida, National Institute of 

Radiological Sciences (Japan) 

7. 40050 – Comparison of Soil-to-plant Transfer Factors for 

Rice and Wheat Grains 

Shigeo Uchida, Keiko Tagami, National Institute of 

Radiological Sciences (Japan) 

SESSION H8: Site Characterization and Modeling of 

Geological Environment (1) 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Chin-Fu Tsang, LBNL (USA) and Motoi 

Kawanishi, CRIEPI (Japan) 

1. 40121 – Development of Characterization Methodology 

for Fault Zone Hydrology 

Kenzi Karasaki, LBNL (USA); Tiemi Onishi US 

Geological Survey (USA); Erika Gasperikova LBNL 

(USA) Junichi Goto, Tadashi Miwa, Hiroyuki Tsuchi, 

NUMO (Japan); Keiichi Ueta, Kenzo Kiho, Kimio 

Miyakawa, CRIEPI (Japan) 

2. 40189 – An attempt to evaluate horizontal crustal 

movement by geodetic and geological approach in the 

Horonobe area, northern Hokkaido, Japan 

Tetsuya Tokiwa, Koichi Asamori, Tadafumi Niizato, 

Tsuyoshi Nohara, JAEA (Japan); Yuki Matsuura, Hitachi 

Zosen Co., Ltd. (Japan); Hideki Kosaka, Kankyo 

Chishitsu Co., Ltd. (Japan) 

3. 40054 – Relationship between hypocentral distribution 

and geological structure in the Horonobe area, northern 

Hokkaido, Japan 

Tetsuya Tokiwa, Koichi Asamori, Naoto Hiraga, Osamu 

Yamada, Hideharu Yokota, JAEA (Japan); Hirokazu 

Moriya, Tohoku University (Japan); Hikaru Hotta, Itaru 

Kitamura, Construction Project Consultants, Inc. (Japan) 

—————— Break ——————— 

4. 40062 – Technical know-how for modeling of geological 

environment (1) Overview and groundwater flow

modeling 

Hiromitsu Saegusa, Shinji Takeuchi, Keisuke Maekawa, 

Hideaki Osawa, Takeshi Semba, JAEA (Japan) 

5. 40066 – Technical Know-how For Modeling Of 

Geological Environment (2) Geological Modeling 

Toshiyuki Matsuoka, Kenji Amano, Hideaki Osawa, 

Takeshi Semba, JAEA (Japan) 

6. 40039 – <strong>The</strong> long-term stability of geological 

environments in the various rock types in Japan from the 

perspective of uranium mineralization 

Eiji Sasao, JAEA (Japan) 

SESSION D5: Measurement and Estimation 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Vladimir Daniska, DECONTA, a.s. 

(Slovakia) and Yukihiko Iguchi, JNES (Japan) 

1. 40045 – Improvement of Radioactivity Inventory 

Evaluation Procedure In Preparatory Tasks for 


Ken-ichi Tanaka, Hidenori Tanabe, Hideaki Ichige, 

JAPC (Japan) 

2. 40202 – Verification of Source Term Analysis System for 

Decommissioning Wastes from CANDU Reactor 

Dong-Keun Cho, Gwang-Min Sun, Jongwon Choi, 

KAERI (Korea Rep.); Donghyeun Hwang, Hak-Soo Kim, 

Tae-Won Hwang, KHNP (Korea Rep.) 

3. 40294 – Evaluation of <strong>The</strong> Activated Radioactivity of 

Turbine Equipments in BWR 

Masato Watanabe, Motonori Nakagami, Chubu Electric 

Power Co., Inc. (Japan) 

—————— Break ——————— 

4. 40014 – Optimization of Quantitative Waste Volume 

Determination Technique for Hanford Waste Tank 

Closure 

Yi Su, David L. Monts, Ping-Rey Jang, Zhiling Long, 

Walter P. Okhuysen, Olin P. Norton, Lawrence L. 

Gresham, Jeffrey S. Lindner, Mississippi State University 

(USA) 

5. 40120 – Implementation of Decommissioning Materials 

Conditional Clearance Process to the OMEGA 

Calculation Code 

Matej Zachar, Vladimir Necas, Slovak University of 

Technology (Slovakia); Vladimir Daniska� DECONTA� 

a.s. (Slovakia) 

6. 40183 – Quantitative determination of the initial 

components in the activated pressure tubes of the 

Wolsong 1st CANDU reactor 

Gwang-Min Sun, Dong-Keun Cho, KAERI (Korea Rep.) 

SESSION R3: ER Techniques 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Takumi Kubota, Kyoto University 

Research Reactor Institute (Japan) and Mark B. Triplett, 

PNNL (USA) 

1. 40286 – Sequential Extraction and Determination of 

Depleted Uranium in the Presence of Natural Uranium in 

Environmental Soil samples by ICP-MS 

Mohamed Amr, Alaa E. Negmeldin, Khalid Al-Saad, A. T. 

Al-Kinani, Qatar University (Qatar); A. I. Helal, Atomic 

Energy Authority (Qatar) 

2. 40096 – Determination of Environmental Uranium 

Concentration by Utilizing Gamma-ray Emission from 

the Progeny Radionuclides 


16 

Tadao Tanaka, Taro Shimada, Takenori Sukegawa, JAEA 

(Japan); Takeshi Ito, Japan ATOX Co., Ltd. (Japan) 

3. 40170 – Natural Radioactivity Levels And Gamma-Ray 

Dose Rate In Soil And Transfer Of Radionuclides From 

Soil To Vegetation Of Some Northern Area Of Pakistan 

Using Gamma-Ray Spectrometry 

Muhammad Ismail, Hasan M.Khan, Peshawar 

University (Pakistan); Khalid Khan, Pinstech Islamabad 

(Pakistan) 

—————— Break ——————— 

4. 40034 – Effect of Fertilizer and Soil Amendments on 

Extraction Yields of Radioiodine and Radiocesium in Soil 

Hirofumi Tsukada, Akira Takeda, Shunichi Hisamatsu, 

Institute for Environmental Sciences (Japan) 

5. 40246 – Impact of Mobile-Immobile Water Domains on 

the Retention of Technetium (Tc-99) in the Vadose Zone 

Danielle Jansik, Dawn Wellman, Elsa Cordova, PNNL 

(USA) 

6. 40207 – Nano Technologies for Environmental 

Remediation 

Harch Gill, Lehigh Nanotech (USA) 

7. 40122 – Remediation of 153Gd-contaminated sand by 

fulvic and humic materials extracted from fallen cherry 

leaves 

Takumi Kubota, Kyoto University Research Reactor 

Institute (Japan) 

SESSION L7: Storage and Disposal Facility 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Sheila M. Brooks, AECL (Canada) and 

TBD 

1. 40095 – Design, Licensing and Commissioning Activities 

for the Waste Storage Facility to Manage the Waste 

Streams from Retubing of Wolsong-1 CANDU 600 MW 

Nuclear Generating Station in South Korea 

Nikos Pontikakis, John Ballantine, Sharon Hugh, Joulien 

Moussalam, AECL (Canada) 

2. 40284 – Microbial Occurrence in Bentonite-based Buffer 

Materials of a Final Disposal Site for Low Level 

Radioactive Waste in Taiwan 

Fong-In Chou, Chia-Chin Li, Tzung-Yuang Chen, 

National Tsing Hua University (Taiwan); Hsiao-Wei Wen, 

National Chung Hsing University (Taiwan) 

3 40153 – Assessing the gas transport mechanisms in the 

Swiss L/ILW concept using numerical modeling 

Irina Gaus, Paul Marschall, Joerg Rueedi, Nagra 

(Switzerland); Rainer Senger, John Ewing, Intera Inc. 

Swiss Branch (Switzerland) 

4. 40283 – <strong>The</strong> progress and results of Demonstration Test 

of Cavern-Type Disposal Facility 

Yoshihiro Akiyama, Kenji Terada, Nobuaki Oda, Tsutomu 

Yada, Takahiro Nakajima, RWMC (Japan) 

SESSION H9: Repository Engineering and 



--------------------------------------------------------------------------------------- 

Session Co-Chairs: Ian G. McKinley, McKinley Consulting 

(Switzerland) and Hidekazu Asano, RWMC (Japan) 

1. 40304 Keynote – Repository engineering and 

demonstration: special challenges for TRU 


2. 40119 – Half-Scale Test: An important step to 

demonstrate the feasibility of the Belgian Supercontainer

concept for disposal of HLW 

Lou Areias, SCK CEN and VUB (Belgium); Bart Craeye, 

Ghent University/Artesis Hogeschool Antwerpen 

(Belgium); Geert De Schutter, Ghent University 

(Belgium); Hughes Van Humbeeck, William Wacquier, 

ONDRAF/NIRAS (Belgium); Alain Van Cotthem, Loic 

Villers Technum-Tractebel (Belgium) 

3. 40175 – Full-Scale Test on Overpack Closure Techniques 

for HLW Repository Operation - Welding Methods and 

UT Systems For Long-Term Structural Integrity of the 

Weld Joint - 

Ario Nakamura, Hidekazu Asano, RWMC (Japan); 

Takashi Furukawa, JAPAN POWER ENGINEERING 

AND INSPECTION CORPORATION (Japan); Kyosuke 

Fujisawa, KOBE STEEL, LTD. (Japan); Susumu 

Kawakami, IHI Corporation (Japan); Takashi Ito, 

Mitsubishi Heavy Industries, Ltd. (Japan) 

4. 40242 – Design Options for HLW Repository Operation 

Technology, (I) Demonstration and Evaluation of 

Remote Handling Technologies 

Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan); 

Hideki Kawamura, Obayashi Corporation (Japan) 

—————— Break ——————— 


Technology, (II) Bentonite Block Forming and Vertical 

Emplacement/JGC 

Hajime Takao, Tatsuhiro Takegahara JGC Corporation 

(Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC 

(Japan) 


Technology, (III) Transportation and Horizontal 

Emplacement of Pre-Fabricated EBS Module (PEM) 

Susumu Kawakami, IHI Corporation (Japan); Hitoshi 

Nakashima, Hidekazu Asano, RWMC (Japan) 


Technology, (IV) Shotclay Technique for Seamless 

Construction of EBS 

Ichizo Kobayashi, Soh Fujisawa, Makoto Nakajima, 

Masaru Toida, Kajima Corporation (Japan); Hitoshi 

Nakashima, Hidekazu Asano, RWMC (Japan) 


Technology, (V) Preliminary Study and Small Scale 

Experiments on the Method of Removal of Buffer 

Material with Salt Solution 

Satohito Toguri, Jiho Jang, Takashi Ishii, Mitsunobu 

Okihara, Kengo Iwasa, SHIMIZU CORPORATION 

(Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC 

(Japan) 

SESSION H10: Site Characterization and Modeling of 

Geological Environment (2) 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Kenji Karasaki, LBNL (USA) and Yuji 

Ijiri, Taisei Corporation (Japan) 

1. 40135 – Dry-run of Site Investigation Planning using the 

Manual for Preliminary Investigation in Japan 

Shigeki Akamura, Tadashi Miwa, NUMO (Japan); 

Tatsuya Tanaka, Obayashi Corporation (Japan); Hiroshi 

Shiratsuchi, Tokyo Electric Power Service Co.,ltd. 

(Japan); Atsushi Horio, DIA CONSULTANTS CO., Ltd. 

(Japan) 

2. 40070 – Evaluation of the long-term evolution of the 

groundwater system in the Mizunami area, Japan 

Takashi Mizuno, Teruki Iwatsuki, JAEA (Japan); Antoni 

E. Milodowski, British Geological Survey (UK) 


17 

3. 40077 – Study on the Estimation Error Caused by Using 

One-dimensional Model for the Evaluation of Dipole 

Tracer Test 

Yuji Ijiri, Yumi Naemura, Taisei Corporation (Japan); 

Kenji Amano, Keisuke Maekawa, Kunio Ota, Takanori 

Kunimaru, Atsushi Sawada, JAEA (Japan) 

4. 40056 – Development of Comprehensive Techniques for 

Coastal Site Characterisation (1) Strategic Overview 

Kunio Ota, Kenji Amano, Tadafumi Niizato, JAEA 

(Japan); W Russell Alexander, Bedrock Geosciences 

(Switzerland); Yoshiaki Yamanaka, Suncoh Consultants 

(Japan) 

—————— Break ——————— 

5. 40052 – Development of comprehensive techniques for 

coastal site characterisation: (3) Conceptualisation of 

long-term geosphere evolution 

Tadafumi Niizato, Kenji Amano, Kunio Ota, Takanori 

Kunimaru, JAEA (Japan); Lanyon Bill, Nagra 

(Switzerland); W Russell Alexander, Bedrock 

Geosciences (Switzerland) 

6. 40048 – Development of Comprehensive Techniques for 

Coastal Site Characterisation (2) Integrated 

Palaeohydrogeological Approach for Development of Site 

Evolution Models 

Kenji Amano, Tadafumi Niizato, Hideharu Yokota, Kunio 

Ota, JAEA (Japan); Bill Lanyon, Nagra (Switzerland); W 

Russell Alexander, Bedrock Geosciences (Switzerland) 

7. 40041 – Development of Methodology of Groundwater 

Flow and Solute Transport Analysis in the Horonobe Area, 

Hokkaido, Japan 

Keisuke Maekawa, Hitoshi Makino, Hiroshi Kurikami, 

Tadafumi Niizato, Manabu Inagaki, Makoto Kawamura, 

JAEA (Japan) 

SESSION D6: Waste Treatment and Non-Reactor 


--------------------------------------------------------------------------------------- 

Session Co-Chairs: Takeshi Ishikura, IAE (Japan) and 

Hitoshi Sakai, Toshiba Corporation (Japan) 

1. 40105 – Estimation of Radioactivity of Graphite Blocks 

in Tokai Power Station using Statistical Method 

Masaaki Nakano, Fuji Electric Holdings Co., Ltd. 

(Japan); Hisashi Mikami, Fuji Electric Systems Co., Ltd. 

(Japan); Hideaki Ichige, Shinich Tsukada, JAPC 

(Japan) 

2. 40115 – <strong>The</strong> treatment of hexavalent chromium in waste 

liquid from Fugen Decommissioning 

Yuki Yahiro, Seiji Yamamoto, Koji Negishi, Hitoshi Sakai, 

Tadashi Fukushima, Norimasa Yoshida, Toshiba 

Corporation (Japan); Nobuo Ishizuka, Yuji Sato, Wataru 

Fujiwara, JAEA (Japan) 

3. 40201 – Characterization of Radioactive Waste from Side 

Structural Components of a CANDU Reactor for 

Decommissioning Applications in Korea 

Rizwan Ahmed, Gyunyoung Heo, Kyung Hee University 

(Korea Rep.); Dong-Keun Cho, Jongwon Choi, KAERI 

(Korea Rep.) 

4. 40068 – Uranium refining and conversion plant 

decommissioning project 

Naoki Zaima, Yasuyuki Morimoto, Noritake Sugitsue, 

Kazumi Kado, JAEA (Japan) 

—————— Break ——————— 

5. 40161 – Radiochemical Characterization of Concrete 

Samples from the Environment of Accelerator Facilities 

Dorothea Schumann, Paul Scherrer Institute 

(Switzerland)

SESSION M2: Public Involvement 


--------------------------------------------------------------------------------------- 

Session Chair: Koji Nagano, CRIEPI (Japan) 

1. 40288 – Activities of the OECD/NEA in the Field of 

Stakeholder Confidence for Radwaste Management and 



2. 40219 – A Comparative Study of Stakeholder 

3. 

Participation in the Cleanup of Radioactive Wastes in the 

US, Japan and UK 

Mito Akiyoshi, Senshu University (Japan); William 

Lawless, Fjorentina Angjellari-Dajci, Paine College 

(USA), Christian Poppeliers, Augusta State University 

(USA) ; John Whitton National Nuclear Laboratory (UK) 

40076 – Territorial Integration of the Geological 

Repository in France 

Gérald Ouzounian, Sebastien Farin, Roberto Miguez, 

Jean-Louis Tison, Andra (France) 

4. 40256 – Public Involvement in the Closure of Fernald 

John Bradburne, Bradburne Consulting LLC (USA) 

POSTER SESSIONS 

Monday PM, October 4, 2010 

Tuesday AM&PM, October 5, 2010 

SESSION L8: L/ILW Poster 

Room 102 

--------------------------------------------------------------------------------------- 

1. 40006 – A GoldSim Modeling Approach to Safety 

Assessment of an LLW Repository System 

Youn Myoung Lee, Jongtae Jeong, Jongwon Choi, 

KAERI (Korea Rep.) 

2. 40011 – Gas Migration Mechanism of Saturated 

Highly-compacted Bentonite and its Modeling 

Yukihisa Tanaka, Michihiko Hironaga, Koji Kudo, 

CRIEPI (Japan) 

3. 40012 – Development of numerical simulation method for 

gas migration through highly-compacted bentonite using 

model of two-phase flow through deformable porous 

media 

Yukihisa Tanaka, CRIEPI (Japan) 

4. 40020 – Planning of Large-scale In-situ Gas Generation 

Experiment in Korean Radioactive Waste Repository, 

Juyoul Kim, Sukhoon Kim, FNC Technology Co. (Korea 

Rep.); Jinbeak Park, Sungjoung Lee, KRMC (Korea Rep.) 

5. 40024 – Estimation and measurement of porosity change 

in cement paste 

Eunyong Lee, Haeryojng Jung, Ki-jung Kwon, KRMC 

(Korea Rep.); Do-Gyeum Kim, Korea Institute of 

Construction Technology (Korea Rep.) 

6. 40082 – Separation and Recovery of Sodium Nitrate from 

Low-level Radioactive Liquid Waste by Electrodialysis 

Yoshihiro Meguro, A Kato, Y Watanabe, Kuniaki 

Takahashi, JAEA (Japan) 

7. 40109 – Study on Mechanical Influence of Gas 

Generation and Migration on Engineered Barrier System 

in adioactive Waste Disposal Facility 

Mamoru Kumagai, JNFL (Japan); Shuichi Yamamoto, 

Kunifumi Takeuchi, Obayashi Corporation (Japan); 

Yukihisa Tanaka, Michihiko Hironaga, CRIEPI (Japan) 

8. 40192 – Liquid Radioactive Wastes in the Republican 


18 

Point of Burial Radioactive Waste of Tajikistan 

Nazirzhon Buriev, Physical-Technical Institute of the 

Academy of Sciences of the Republic of Tajikistan 

(Tajikistan) 

9. 40197 – Inspection of Burial Points of Radioactive Waste 

in the Republic of Tajikistan� 

Nazirzhon Buriev, Physical-Technical Institute of the 

Academy of Sciences (Tajikistan); Akram Juraev, 

Dzhamshed Abdushukurov, S.U.Umarov Physical 

Technical Institute of Academy of Sciences of Republic of 

Tajikistan, (Tajikistan) 

10. 40198 – Pre-treatment Method of Powder Waste for 

Vitrification 

Deun-Man Kim, Hyun-Je Cho, Jong-Kil Park, Nuclear 

Engineering & Technology Institute (Korea Rep.) 

11. 40208 – Characterization and Pre-treatment of LLW in 

Turkey 

Ahmet Erdal Osmanlioglu, Turkish Atomic Energy 

Authority (Turkey) 

12. 40221– Decontamination of Radioactive Concrete Wastes 

by <strong>The</strong>rmal and Mechanical Processes 

B.Y. Min, Ki Won Lee, Un Soo Chung, KAERI. W.K Choi, 

Kune-Woo Lee, KAERI (Korea Rep.) 

13. 40302 – Latex Particles Functionalized with Transition 

Metals Ferrocyanides for Cesium Uptake and 

Decontamination of Solid Bulk Materials 

Dmitry Marinin, Valentin Avramenko, Dmitry Marinin, 

Valentin Sergienko, Institute of Chemistry FEDRAS 

(Russia); Veniamin Zheleznov, Irina Sheveleva, Russian 

Academy of Sciences (Russia) 

SESSION H11: SF/TRU/HLW Poster 

Room 102 

--------------------------------------------------------------------------------------- 

1. 40001 – Investigation of Colloid-facilitated Effects on the 

Radionuclides Migration in the Fractured Rock with a 

Kinetic Solubility-Limited Dissolution Model 

Chun-Ping Jen, Nation Chung Cheng University 

(Taiwan); Neng-Chuan Tien, Energy and Environment 

Research� Laboratories, Industrial Technology Research 

Institute (Taiwan) 

2. 40013 – Modeling hydraulic conductivity and swelling 

pressure of several kinds of bentonites affected by salinity 

of water 

Yukihisa Tanaka, Takuma Hasegawa, Kunihiko 

Nakamura, CRIEPI (Japan) 

3. 40018 – Current R & D activities of the study on 

long-term stability of geological environments 

Takahiro Hanamuro, Kenichi Yasue, Yoko 

Saito-Kokubu,Koichi Asamori, Tsuneari Ishimaru, Koji 

Umeda, JAEA (Japan) 

4. 40019 – In Situ Stress Measurements in Siliceous 

Mudstones at Horonobe Underground Research 

Laboratory, Japan 

Hiroyuki Sanada, Takahiro Nakamura, Yutaka Sugita, 

JAEA (Japan) 

5. 40038 – Low alkaline Cement Used in the Construction 

of a Gallery in the Horonobe Underground Research 

Laboratory 

Masashi Nakayama, Haruo Sato, Yutaka Sugita, Seiji Ito, 

JAEA (Japan); Masashi Minamide, Yoshito Kitagawa, 

Taisei Corporation (Japan) 

6. 40040 – Effects of Nitrate on Nuclide Solubility for 

Co-location Disposal of TRU Waste and HLW 

Gento Kamei, Morihiro Mihara, JAEA (Japan); 

Toshiyuki Nakazawa, Norikazu Yamada, Mitsubishi

Materials Corp. (Japan) 

7. 40047 – A study on groundwater infiltration in the 


Hideharu Yokota, Yamamoto Yoichi, Keisuke Maekawa, 

Minoru Hara, JAEA (Japan) 

8. 40051 – Effective Use of Uranium Resources and 

Dissolution of Recovery Uranium Storage Accumulation 

by a Uranium Multi-recycle System 

Yuzo Yamashita, Yuzo Yamashita, Takeshi Nakamura, 

Kyushu University (Japan) 

9. 40053 – Advanced ORIENT Cycle - Progress on Fission 

Product Separation and Utilization 

Isao Yamagishi, Masaki Ozawa, JAEA (Japan); Hitoshi 

Mimura, Tohoku Univ.(Japan); Shohei Kanamura, Koji 

Mizuguchi, Toshiba Corporation (Japan) 

10. 40064 – Hydrogeological Characterization based on Long 

Term Groundwater Pressure Monitoring 

Shuji Daimaru, Ryuji Takeuchi, Masaki Takeda, 

Masayuki Ishibashi, JAEA (Japan) 

11. 40065 – Development Of <strong>The</strong> Quality Management 

System For Borehole Investigations: (2) Quality 

Management Systems For Hydrochemical Investigations, 

Takanori Kunimaru, Kunio Ota, Kenji Amano, JAEA 

(Japan); W Russell Alexander, Bedrock Geosciences 

(Switzerland) 

12. 40067 – An Analytical Model on the Sealing Performance 

of Space for the Design of Buffer Material and Backfill 

Material 

Haruo Sato, JAEA (Japan) 

13. 40069 – Current Status of Horonobe URL Project in 

Construction Phase 

Hironobu Abe, Koichiro Hatanaka, JAEA (Japan) 

14. 40074 – Development of New Ultrafiltration Techniques 

Maintaining In-Situ Hydrochemical Conditions for 

Colloidal Study 

Daisuke Aosai, Yuhei Yamamoto, Takashi Mizuno, JAEA 

(Japan) 

15. 40089 – Sorption Behavior of Iodine on Calcium Silicate 

Hydrates Formed as a Secondary Mineral 

Keisuke Shirai, Yuichi Niibori, Akira Kirishima, Hitoshi 

Mimura, Tohoku Univ. (Japan) 

16. 40098 – Development of the Quality Management System 

for Borehole Investigations: (1) Quality Management 

System for Hydraulic Packer Tests 

Shinji Takeuchi, Takanori Kunimaru, Kunio Ota, JAEA 

(Japan); Bernd Frieg, Nagra (Switzzerland) 

17. 40101 – Comparison of Post-Irradiation Experimental 

Data and <strong>The</strong>oretical Calculations for Inventory 

Estimation of Long-Lived fission Products in Spent 

Nuclear Fuel 

Shiho Asai ,Yukiko Hanzawa, Keisuke Okumura, Hideya 

Suzuki, Masaaki Toshimitsu, Nobuo Shinohara, JAEA 

(Japan); Kensuke Suzuki, Satoru Kaneko, Tokyo Electric 

Power Company (Japan) 

18. 40103 – Selective Uptake of Palladium from High-Level 

Liquid Wastes by Hybrid Microcapsules 

Hitoshi Mimura, Takashi Sakakibara, Yan Wu, Yuichi 

Niibori, Tohoku Univ. (Japan); Shin-ichi Koyama, 

Takashi Ohnishi, JAEA (Japan) 

19. 40113 – An Empirical Model To Determine <strong>The</strong> Modes 

Of Corrosion Of Carbon Steel 

Toshikatsu Maeda, Masatoshi Watanabe, Seiji Takeda, 

Shinichi Nakayama, JAEA (Japan) 

20. 40124 – Trends in Scenario Development Methodologies 

and Integration in NUMO's Approach 

Takeshi Ebashi, Katsuhiko Ishiguro, NUMO (Japan); 

Keiichiro Wakasugi, JAEA (Japan); Hideki Kawamura, 


19 

Obayashi Corporation (Japan); Irina Gaus, Stratis 

Vomvoris, Andrew Martin, Nagra (Switzerland); Paul 

Smith, Safety Assessment Management (UK) 

21. 40137 – Development of Methodology to Construct a 

Generic Conceptual Model of River-valley Evolution for 

Performance Assessment of HLW Geological Disposal 

Makoto Kawamura, Kenichi Yasue, Tadafumi Niizato, 

Shin-ichi Tanikawa, JAEA (Japan) 

22. 40162 – Long-term Stability of Bentonite Material Used 

as Engineered Barrier for Radioactive Waste Disposal , 

Vera Jedinakova-Krizova, Katerina Videnska, Institute of 

Chemical Chechnology Prague (Czech Republic); Eduard 

Hanslik, T.G.M. Water Research Institute (Czech 

Republic) 

23. 40173 – Development of a methodology for evaluating 

the performance of host rock for geological disposal 

based on surface-based investigations 

Manabu Inagaki, Keisuke Maekawa, JAEA (Japan); 

Tatsuya Tanaka, Shuji Hashimoto, Obayashi Corporation 

(Japan) 

24. 40176 – Structural Integrity Evaluation Approach for 

PWR Spent Nuclear Fuel 

Yun Seog Nam, Seong Ki Lee, Yong Hwan Kim, Jeon 

Kyeong Lak, Choi Ki Sung, Chang Sok Cho, KNF (Korea 

Rep.) 

25. 40200 – Development of Program Categories to Assess 

the Radiological Dosage during Spent Fuel Transportation 

Suhong Lee, Sangwon Shin, Enesys, Jaemin Lee, 

Enesys.Co., Ltd. (Korea Rep.); Kiyroul Seong, 

Jeonghyoun Yoon, KRMC (Korea Rep.) 

26. 40225 – Exploiting synergies between the UK & Japanese 

geological disposal programmes� 

Ellie Scourse, Atkins (UK); Hideki Kawamura, Obayashi 

Corporation (Japan); Ian G. McKinley, McKinley 

Consulting (Switzerland) 

27. 40239 – Realistic Consequence Analysis of River Erosion 

Scenarios for a HLW epository� 

Kaname Miyahara, Makoto Kawamura, Manabu Inagaki, 

JAEA (Japan); Ian G. McKinley, McKinley Consulting 

(Switzerland); Michael J. Apted, Monitor Scientific LLC 

(USA) 

28. 40272 – Removal of Fission Products in the Spent 

Electrolyte Using Iron Phosphate Glass as a Sorbent 

Ippei Amamoto, Masami Nakada, Yoshihiro Okamoto, 

JAEA (Japan); Naoki Mitamura, Tatsuya Tsuzuki, Central 

Glass Co., Ltd. (Japan); Yasushi Takasaki, Atsushi 

Shibayama, Akita University (Japan); Tetsuji Yano, Tokyo 

Institute of Technology (Japan) 

29. 40295 – Propagation and Interactions of Acoustic Waves 

in a Waveguide Attached at the Surface of Rock 

Jin-Seop Kim, Kyung-Soo Lee, S. Kwon, KAERI (Korea 

Rep.); Gye-Chun Cho, KAIST (Korea Rep.) 

SESSION D7: D&D Poster 

Room 102 

--------------------------------------------------------------------------------------- 

1. 40009 – Decommissioning of the KRR-1 & 2 Research 

Reactors at KAERI; Summary on the Project 

Jin Ho Park, KAERI (Korea Rep.) 

2. 40075 – Methods of Selected Input Calculation Data 

Verification And <strong>The</strong>ir Influence On Decommissioning 

Cost In the OMEGA Code 

Frantisek Ondra. Vladimir Daniska, Ivan Rehak, Oto 

Schultz, DECONTA, a.s. (Slovakia); Vladimir Necas, 

Slovak University of Technology in Bratislava (Slovakia) 

3. 40078 – Experience of MR and RFT Reactors’

Decommissioning in RRC “Kurchatov Institute” 

Alexander V. Chesnokov, Victor Volkov, Sergey 

Semenov, Vitaly Pavlenko, Vyacheslav Kolyadin, D. 

Muzrukova, RRC "Kurchatov Institute" (Russia); Artur 

Arustamov, SUE SIA "Radon" (Russia) 

4. 40126 – Detailed Standardized Decommissioning 

Parameters Calculation for Larger echnological 

Aggregates and Relevant Buildings in Nuclear Power 

Plants using the OMEGA Code 

Peter Bezak, Vladimir Daniska, DECONTA, a.s. 

(Slovakia); Ivan Rehak, DECOM, a.s . (Slovakia) 

5. 40190 – Dismantling Method of Fuel Cycle Facilities 

Obtained by Dismantling of the JRTF 

Fumihiko Kanayama, JAEA (Japan) 

6. 40191 – Computer Simulation of Cryogenic Jet Cutting 

for Dismantling Highly Activated Facilities 

Sung-Kyun Kim, Kune-Woo Lee, KAERI (Korea Rep.) 

7. 40193 – Strippable Core-shell Polymer Emulsion for 

Decontamination of Radioactive Surface Contamination 

Bum-Kyoung Seo, Bum-Kyoung Seo, Kune-Woo Lee, 

KAERI (Korea Rep.) 

SESSION R4: ER Poster 

Room 102 

--------------------------------------------------------------------------------------- 

1. 40025 – Improvement of quicklime mixing treatment by 

carbon dioxide ventilation 

Yuki Nakagawa, Hisayoshi Hashimoto, Hitachi 

Construction Machinery Co., Ltd. (Japan); Koichi Suto, 

Chihiro Inoue, Tohoku University (Japan) 

2. 40166 – Procedure and result of decommission of R&D 

facility of Uranium fuel 

Hirokazu Tanaka, Masao Shimizu, Ryoji Tanimoto, 

Kazuhiko Maekawa, Shinzo Ueta, Mitsubishi Materials 

Corporation (Japan); Susumu Tojo, SERNUC 

Corporation (Japan) 

3. 40210 – Photo-Catalytic Degradation of Industrial 

Pollutant by Using Nano-technology and Solar energy in 

the Eastern Area of Kingdom of Saudi Arabia 

Osama S.Y. Mohamed, Abdullah A. Al Jaafari, King 

Faisal University (Saudi Arabia) 

4. 40224 – Hydrogen Production from a PV/PEM 

Electrolyzer System Using a Neural-Network-Based 

MPPT Algorithm 

Abd El-Shafy Nafeh, Electronics Research Institute 

(Egypt) 

5. 40301 – Cement based solidification / stabilization of 

industrial contaminated soil using various cement 

additives 

Grega E. Voglar, RDA (Slovenia); D. Lestan, 

Agronomy Department, Biotechnical Faculty, University 

of Ljubljana, (Slovenia) 

SESSION M3: EM/PI Poster 

Room 102 

--------------------------------------------------------------------------------------- 

1. 40099 – Removal of Fluorine and Boron from 

Groundwater Using Radiation-induced Graft 

Polymerization Adsorbent at Mizunami Underground 

Research Laboratory 

Yosuke Iyatomi, Hiroyuki Hoshina, Noriaki Seko, Noboru 

Kasai, Yuji Ueki, Masao Tamada, JAEA (Japan) 

2. 40184 – <strong>The</strong> Optimized Risk Management of the Waste 

from NORM and Nuclear Industries - How to Harmonize 


20 

Risk from Various Sources 

Yoko Fujikawa, Kyoto University Research Reactor 

Institute (Japan); Michikuni Shimo, Fujita Health 

University (Japan); Hidenori Yonehara, National 

Instuitute of Radiological Sciences (Japan); Tadashi 

Tujimoto, Electron Science Institute (Japan) 

3. 40205 – Exposure Dose Evaluation of Worker at 

Radioactive Waste Incineration Facility on KEARI 

Sang Kyu Park, Jong Seon Jeon, Younhwa Kim, Jaemin 

Lee, NESYS.CO., LTD (Korea Rep.); Ki Won Lee, KAERI, 

(Korea Rep.) 

4. 40209 – Scenario Development for Safety Assessment of 

Waste Repository for Feasibility Study on Transmutation 

of Spent Nuclear Fuel into LILW using PEACER 

Sung-yeop Kim, Kun Jai Lee, KAIST (Korea Rep.)

List of Exhibitors 

<strong>The</strong> following exhibitors are greatly appreciated: 

(Exhibitors in Alphabetical Order) 

�� Belgoprocess 

�� Central Research Institute of Electric Power Industry 

�� Fuji Electric Systems Co., Ltd. 

�� GEOSCIENCE RESEARCH LABORATORY Co., Ltd. 

�� JAPAN NUCLEAR FUEL LIMITED 

�� Kajima Corporation 

�� Kobe Steel, Ltd. (KOBELCO) 

�� Mitsubishi Materials Corporation 

�� NIPPON KOEI Co., Ltd. 

�� Nuclear Waste Management Organization of Japan (NUMO) 

�� OBAYASHI CORPORATION 

�� PIERCAN USA, Inc. 

�� Raax Co., Ltd. / Earth Scanning Association 

�� SHIMIZU CORPORATION 

�� TAISEI CORPORATION 

�� Toshiba Corporation 

�� Web I Laboratories, Inc. 

�� Westinghouse Electric Company 

21

Nuclear Power Generation / 

Nuclear Fuel Cycle Related Facilities 

1 

Fuji�s 3 unique technologies which include remote handling, radwaste 

treatment and high temperature gas-cooled reactor are contributing for 

ensuring energy resources for longer than a century. 

Remote Handling & Fuel Fabrication Technologies 

Fuel Pellets Manufacturing Facility Internal Equipment within the Glove Box 

Performance of 

volume reduction 

(ion-exchange resin) 

(1/10–1/20) 

Exterior View of Fuji Resin Reducer(LPOP) 

(LPOP : Low Pressure Oxidation Process) 

Commercial High 

Temperature Gascooled 

Reactor 

(HTGR) 

Active Core 

(Fuel Blocks) 

Core Bottom 

Structure 

pellet 

inspection 

(<strong>The</strong> HTGR heat utilization plant) 

Contact : Fuji Electric Corp. of America 

Phone 201-712-0555 FAX 201-368-8258 

URL http//www.fujielectric.com/fecoa / 

Gripper 

: Fuji Electric Systems Co., Ltd. 

Phone +81-44-329-2169 FAX +81-44-329-2178 

URL http://www.fujielectric.com/ 

�� 

Fuel Transfer System 

2 Radwaste Treatment Systems & Decommissioning Technologies 

Cutting Torch 

Reactor Vessel 

Verification test of remote dismantling system 

(Tokai-1 Decommissioning) 

3 Reactor Technology for High Temperature Gas-cooled Reactor 

24 

HTTR Core internals 

Top view of the core 

(Outer diameter 4.25m) 

(HTTR : High Temperature Engineering Test Reactor)

Abstracts 

OPENING SESSION 

“Nuclear Energy Strategy for Sustainable Growth: 

Aiming at Green Innovation and Life Innovation” 

Tatsujiro Suzuki, Vice Chairman, Japan Atomic Energy Commission (Japan) 

Japan Atomic Energy Commission released a report, entitled “Nuclear Energy Strategy for Growth” on May 25, 

2010. <strong>The</strong> report is intended to contribute the government’s “New Growth Strategy” which was eventually released in 

June 18, 2010. This JAEC’s report emphasizes the role of nuclear energy technologies for “green innovation” and “life 

innovation” which are two primary components of New Growth Strategy. <strong>The</strong> key messages included in the report are 

“strengthening capability to meet global challenges” and “enhancing citizen’s confidence in nuclear energy 

technologies”. 

<strong>The</strong> report listed five major recommendations as follows. 

1. Improve performance of nuclear power plants aim 

2. ing at the world best class and take necessary measures for adding more nuclear plants to achieve 

increased share of nuclear power in total power generation for “green innovation” Promote industrial 

application of radiation technologies, such as medial, agricultural and scientific applications, as 

“strategic industry” for life innovation 

3. Develop social and economic environments for achieving above goals, which include; (1) improve 

scientific literacy (2) take new initiatives for data/information disclosure (3) enhance economic 

visualization of CO2 merits of nuclear energy (4) reform safety regulatory structure to improve 

public confidence (5) develop better approaches for improving local community development and 

nuclear related facilities (6) develop public/private human network in Asia. 

4. Develop appropriate measures to meet increasing global needs for nuclear power and radiation 

technologies 

5. Develop long term social platform for sustainable growth for nuclear energy technologies. 

“U.S. Office of Environmental Management - World Leaders in Nuclear Cleanup and Construction” 

Dae Y. Chung, Principal Deputy Assistant Secretary for Environmental Management, US DOE (USA) 

U.S. Department of Energy Principal Deputy Assistant Secretary for Environmental Management, Dae Chung, will 

present an overview of the progress, challenges, and future opportunities associated with the world's largest nuclear 

cleanup program. <strong>The</strong> Office of Environmental Management (EM) is responsible for the safe cleanup of over two 

million acres of land located in 34 states. EM employs more than 30,000 people – primarily scientists, engineers, and 

hazardous waste technicians, and has an annual budget of approximately $6 billion. 

“Recent Progress in Radioactive Waste Management in Korea” 

Ho Taek Yoon, Senior Vice President, KRMC (Korea Rep.) 

Republic of Korea is set to become a major nuclear power generation country. Nuclear energy initiative is a 

strategic priority for R.O.K, and of which nuclear share in electricity generation is planned to become 56% by 2030, with 

nuclear installed capacity of 35 GWe. Currently, 20 reactors with installed capacity of 17.7 GWe provide almost 40% of 

domestic electricity demand. However, the very active nuclear power program inevitably causes increase in the build-up 

of the radioactive wastes including Low- and Intermediate-Level radioactive Waste (LILW) and spent nuclear fuel. 

<strong>The</strong>refore, reliable and effective management of the radioactive waste and spent nuclear fuel has become a key agenda 

for the continuous growth of the nuclear power program. 

LILW generated by the nuclear power plants is being tentatively stored at each reactor site. <strong>The</strong> accumulated 

amount of LILW in Korea’s nuclear power plants as of Dec. 2009 is approximately 87,000 drums (200 liter drum). And 

together with those of research institutes, nuclear fuel manufacturer, general industry and medical sectors reached 

approximately 112,000 drums (200 liter drum). <strong>The</strong> construction of a final repository is urgently needed to manage the 

radioactive waste effectively and safely. 

Efforts to secure site for the radioactive waste disposal facility in Korea began in 1986. In 18 years from 1986 to 

2004, nine attempts of site selection had been executed unfruitfully. In the 10th attempt which was made in 2005, Korean 

government requested that the public acceptance shall be confirmed through resident voting at the each local government 

volunteering for the repository hosting and decided that the spent nuclear fuel site will be separated from LILW disposal 

site, and the community support packages will be guaranteed by the law. Under the above mentioned basic principles, the 

resident vote was implemented in the four volunteered local governments including Gunsan-kun, Youngdok-kun, Pohang 

city, and Gyeongju city. Among these local governments, Gyeongju city recorded the highest consent vote rate (89.5%) 

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and became the hosting city of LILW repository site. 

After the vote, on January 2, 2006, the Ministry of Knowledge and Economy (MKE) designated Bonggil-ri, 

Yangbuk-myeon, Gyeongju-city, North Gyeongsang Province (approximately 2,100,000 m2) as the LILW repository site. 

<strong>The</strong> first stage Wolsong disposal facility construction will be completed by December 2012 as underground silos type 

with a disposal capacity of 100,000 drums. 

Spent nuclear fuels generated from nuclear power plants are stored at reactor sites pending construction of an 

interim storage facility or final repository. <strong>The</strong> cumulative amount of spent fuel is about 10,761 tons as of Dec. 2009, and 

is expected to increase up to 20,000 tons by 2020. <strong>The</strong> existing storage capacity as of Dec. 2009 is 13,532 tons. 

National policy for spent nuclear fuel management shall be decided in a timely manner through public participation 

process taking into consideration of current international trends and technology development. KRMC launched the spent 

fuel management alternative study and establishment of road map in Dec. 2009 to promote expert group’s consensus and 

for the public involvement in the future which is approaching imminently. 

“Radioactive waste management – achievements, needs and future expectations” 

Irena Mele, Head of Waste Technology Section, Division of Nuclear Fuel Cycle and Waste Technology, IAEA 

<strong>International</strong> nuclear community is facing a period of dynamic changes. <strong>The</strong> growing interest for nuclear energy 

brings new optimism to the community but also new challenges and increased responsibilities. Radioactive waste 

management remains on the agenda even in most developed nuclear countries. Although significant progress has been 

achieved in this area and many disposal solutions for different types of waste have been successfully implemented 

worldwide, unresolved disposal of HLW and spent nuclear fuel remains the major concern of the public related to the use 

of nuclear energy and may become an obstacle for the planned expansion of their nuclear programmes. <strong>The</strong>re are also 

countries that are still missing necessary radioactive waste management infrastructure and, thus, capacity and capability 

to safely manage their low level and intermediate level radioactive waste. In spite of well developed and available waste 

management technologies and good practices many our Members States are still meeting great deficiency in radioactive 

waste management infrastructure and need strong support and assistance to develop adequate level for safe and efficient 

management of their radioactive waste. In particular countries with only institutional waste experience great difficulties 

in providing waste management facilities due to the lack of capacity and expertise. 

Many new countries have also expressed their interest and intention to embark on nuclear energy. When planning 

and preparing for new nuclear build, major attention is given to the construction and operation of nuclear power plant and 

infrastructure building. Much less attention of those countries and also nuclear industry is given to the fact that in parallel 

with the NPP the radioactive waste management infrastructure needs to be developed as well. A big challenge is to 

increase awareness among those countries and also among nuclear industry on burning need to include also preparations 

for spent fuel and radioactive waste management in early planning process. 

Decommissioning of nuclear facilities has become a mature technology and planning for decommissioning and 

implementing a decommissioning plan is commonly accepted practice in most of nuclear countries. However, additional 

encouragement and support is still needed in countries with less experience and smaller programmes. Another 

important issue is cleaning-up of sites with legacy waste emanating from past research, medical or military activities. 

Important legacies from earlier uranium mining are known in many parts of the world. Remediation of land contaminated 

by radioactive material residues is needed in many countries, like in the Central Asian republics, where the remedial 

challenges will require a consolidated international support approach. 

<strong>The</strong> paper will provide an overview of current situation and future prospects in all these areas of radioactive waste 

management, decommissioning and environmental remediation and highlight the role and efforts of the IAEA to ensure 

that adequate safety standards in this area are applied worldwide, that experiences and good practices gained in 

radioactive waste management area are shared and communicated between the countries and potential users, and that 

proper assistance is provided to countries considering nuclear power. 

SESSION L1: Waste Management 

1) 40081 – Radioactive Waste: Feedback of 40-year Operations in France 

Michel Dutzer, Gérald Ouzounian, Roberto Miguez, Jean-Louis Tison, Andra (France) 

France's experience in the management of radioactive waste is supported by forty years of operational activities in 

the field of surface disposal. This feedback is related to three disposal facilities: 

- Centre de la Manche disposal, not far away Cherbourg, from design to post-closure facility, 

- Centre at Soulaines-Dhuys from site selection to design to operation during nearly 20 years, 

- Centre at Morvilliers from site selection to operation for seven years now. 

During the operational period of Centre de la Manche disposal facility (1969-1994), the safety concept for low-and 

intermediate level short lived waste (LIL-SLW) was developed and progressively incorporated in the procedures of the 

facility. <strong>The</strong> facility entered its institutional control period and the experience of this facility has been useful for the 

operating facilities. 

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Centre de l’Aube that took over Centre de la Manche, and Morvilliers for very low level wastes. Both facilities 

currently accommodate the major part of the volume of radioactive wastes that are generated in France. However 

disposal facilities have to be considered as rare resources. <strong>The</strong>n new waste management options are being investigated as 

the disposal of large components or recycling metallic wastes within the nuclear industry. 

2) 40227 – ILW Characterisation and Strategy Development at Chapelcross Power Station, Dumfriesshire, 

Scotland 

Rob Hodgson, Alastair Stewart, Greg Wotherspoon, Magnox North (UK); 

Vince Cane, Dr Rob Thied, Nuclear Technologies PLC (UK) 

<strong>The</strong> decommissioning strategy team at Chapelcross is conducting work to establish a robust and compliant strategy 

for the discharge of its intermediate level waste (ILW) liabilities. 

<strong>The</strong> work is supported by staff from the Environment and Radioactive Waste Management Team at Nuclear 

Technologies PLC. A circa. £1.2M ($1.85M) 32 month contract was placed in 2008 for the development of a fully 

developed strategy and outline concept design, supported by an integrated programme of waste characterisation tasks. 

<strong>The</strong> characterisation programme was designed to improve understanding of the relevant waste streams and to 

underpin future design development by removing risk and uncertainty at the outset of the project. 

<strong>The</strong> characterisation work involved; 

• Desktop studies 

• Trace element analysis and activation calculations, 

• In-situ underwater radiometric surveys 

• Pond sludge sampling and analysis 

• Pond Zeolite skip sampling and analysis 

• Reactor humidrier desiccant sampling and analysis 

In what is believed to represent a worldwide first, in application to intermediate level waste , the project team 

utilised the US EPA’s ‘Data Quality Objectives’ methodology (DQO) to ensure a right first time approach to the delivery 

of the characterisation work. <strong>The</strong> DQO approach has been designed to ensure that, prior to conducting R&D tasks, the 

specification of the task is robustly considered and a comprehensive plan put in place. This approach aligns well with 

project management best practice. DQO will also ensure that specifications for analysis of samples are based upon a clear 

understanding of the future use of the data. 

<strong>The</strong> work was supported by a number of specialist sub-contractors and the entire task managed under the auspices 

of an integrated project team, with stakeholders from all parties 

involved in the key elements of decision making. DQO workshops were run to achieve this consensus approach and 

as a result a successful delivery programme has been implemented. 

<strong>The</strong> paper to be presented will focus on the application of DQO and will demonstrate how the process was used to 

design the various R&D programmes. 

Secondary benefits of the process include a fully transparent audit trail which will allow future analysis of the 

project to track key decision points, and a comprehensive quality/knowledge management system with literally thousands 

of quality records tracing every step of the R&D tasks from inception through to delivery of final reports, thereby 

enabling full consideration of results in future design decisions. 

3) 40226 – <strong>The</strong> Ethics of the Management of Low and Intermediate Radioactive Wastes Generated by 

Cernavoda NPP, a Challenge for the Romanian Specialists 

Gheorghe Barariu, Subsidiary of Technology and Engineering for Nuclear Objectives (Romania) 

In Romania there was an extensive interest on treatment of radioactive wastes generated by nuclear technology used 

in industrial and research field. Still the existence of NPP Cernavoda and the provision for a new nuclear power plant 

impose the existence of the disposal capacities for future radioactive wastes that will be further generated by operation 

and decommissioning of new units. <strong>The</strong> key aspects of the radioactive wastes management at the Romanian Cernavoda 

NPP equipped with CANDU 600 reactors result from missing of a Radioactive Waste Treatment Plant. <strong>The</strong> Strategy for 

radioactive waste management was elaborated by the former National Agency for Radioactive Waste (ANDRAD), the 

jurisdictional authority for final disposal and the coordination of nuclear spent fuel and radioactive waste management 

(Order 844/2004), with attributions established by Governmental Decision (GO) 31/2006. <strong>The</strong> Strategy specifies the 

commissioning of the Saligny L/IL Radwaste Repository near Cernavoda NPP, in 2014. <strong>The</strong> new Agency AN&DR, 

Nuclear Agency and for Radioactive Waste which was appointed in 2010, probable will follows the same Strategy. A lot 

of constraints, including limited available surface and bearing capacity on site of the Saligny Repository near Cernavoda 

NPP limits the possibilities for selecting of the technologies for radioactive wastes treatment. For the new NPP not sised 

and located yet wil be provided a separate strategy. Based on the input data the main constraints for the new repository 

design were identified: Cernavoda NPP specific waste characterization is not finalized, the wastes are not properly 

conditioned for the final disposal, environmental constraints due to characteristics of the Saligny Repository, regulatory 

constraints, public acceptance constraints, constraints related Utility concept for treatment of radioactive waste, 

uncertenties related the New Governamental Agency AN&DR. For every constraint mentioned above, suitable measures 

will be taken to reduce the uncertainties. <strong>The</strong> most important technological dilema is related to selection betwen 

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technogies that implies impact on present generation ( incineration, radwaste transfer from ss drums to cs drums and ss 

drums supercompaction ), and technologies that confined the tritium and C-14 in the Repository with impact for next 

generations. <strong>The</strong> paper will present the implicative of this selection on the future Radioactive Waste Treatment Plant. 

4) 40149 – CEA's radioactive waste and unused fuel inventory - Marcoule site example 

Jean-Guy Nokhamzon, Marc Butez, Ddaniel Fulleringer, CEA (France) 

In the field of clean up and decommissioning of nuclear facilities the knowledge of the radioactive waste and 

unused fuel inventories is crucial. For CEA (French Commission in charge of Atomic Energy and Alternative Energies), 

this knowledge is a key issue : in the short term for the flows stream lining as well as in the middle and long terms for 

investment strategy in nuclear support facilities. 

<strong>The</strong>se needs appear in a larger extent at the national scale in France, with the enforcement of two laws promulgated 

in 2006 dealing with radioactive waste and with transparency in Nuclear Safety. 

More generally, a precise and reliable knowledge of inventory and flow production forecast is indispensable to give 

pertinent answers to the questions related to waste valuation. Thus, all these subjects are also worked out with our 

authorities in charge of financial aspects because there are essential for the cost control of clean-up and decommissioning 

projects CEA has to manage. In effect, in these projects, waste management represents a great part of cost and risk. 

For all these reasons CEA has carried out an important work to collect all his waste and unsued fuel inventories. For 

this reason we have developed a new "INFLUVAL" data base which will be combined with a softare featured with 

coherency checking and analysis modules. Indeed, we must guarantee uniqueness and relevance of data coming from 

decommissioning projects and facility managers. 

As a consequence, in 2009 CEA Clean-up and Decommissioning Division has decided to perform, , a 

comprehensive inventory of waste and unused fuel with the following objectives : 

- Consolidation of stock and expected flows, 

- Evaluation of reference costs 

- Help to dimensioning and investment decision for facilities programs ( storage, waste conditioning facility,…) 

- Validation of waste cost ratio. 

In this paper we will take Marcoule as a more precise example. On this CEA center, heavy operations of clean-up 

and decommissioning are carried out on the first French reprocessing plants –UP1-, and the related workshops. 

<strong>The</strong>ses installations have been shut down in 1996. Since this date the main operations were focussed on definitive 

shutdown and dismantling of equipments which were to be realized in priority in the goal of reducing radiological 

activity and operating costs. 

Thus, the consolidation of the inventory data is essential to valid a scenario and then to design facilities for the 

following stages of the UP1 project. 

5) 40258 – Management of historical waste legacy at NRG Petten� 

Renate de Vos, Nuclear Research and consultancy Group (Netherlands) 

Since the 1960s, the Waste Storage Facility (WSF) in Petten has been used to store radioactive waste. In 1993, the 

Dutch government decided that all nuclear waste should be stored in the central storage facility COVRA (Dutch Central 

Organisation for Radioactive Waste). This means that the drums with historical waste must all be opened, their contents 

sorted according to intermediate- and low-level radioactivity and subsequently packed for transport to the COVRA 

facility. 

<strong>The</strong> intermediate-level radioactive waste processing Unit (HAVA-VU) has been designed specifically for this 

purpose. 

In this paper an overview will be given of the process of this HAVA-VU project in the last decade. Subsequently 

the current status and path forward will be described. Also the chosen solutions for waste processing and characterization 

will be discussed. 

6) 40031 – Norwegian Support For Regulations Of Radioactive Waste Management From Uranium Mining 

And Mill Tailings In Central� Asia 

Tamara Zhunussova, Malgorzata Sneve, Astrid Liland, Norwegian Radiation Protection Authority (Norway); 

Alexander Kim, Kazakhstan Atomic Energy Committee (Kazakhstan); 

Ulmas Mirsaidov, Tajikistan Nuclear and Radiation Safety Agency (Tajikistan); 

Baigabyl Tolongutov, Chui Ecological Laboratory of Kyrgiz Republic (Kyrgiz); 

Per Strand, Norwegian Radiation Protection Authority, University of Life Sciences (UMB) (Norway) 

In Central Asia the radioactive waste comes mainly from uranium mining and milling, nuclear weapon testing and 

nuclear power development and other ionizing sources. This waste was produced, to a greater extent, by the 

military-industrial complex and the uranium and non-uranium industry, and, to a lesser extent, by the nuclear industry 

and in the process of use of isotope products. Exploitation and mining of uranium and thorium deposits produce a large 

amount of solid and liquid radioactive waste, as well volatile contaminants which need a proper management. In Central 

Asia the wastes are mainly stored at the surface in large piles and represent a long-term potential health and 

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environmental hazard. <strong>The</strong> process of remediating legacy sites of the past and reducing the threats is now getting under 

way, with the design and implementation of remediation activities, partly with international support. However, there is a 

significant lack in the regulatory basis for carrying out such remediation work, including a lack of relevant radiation and 

environmental safety norms and standards, licensing procedures and requirements for monitoring etc., as well as 

expertise to transform such a basis into practice. Accordingly, the objective of the proposed project is to assist the 

relevant regulatory authorities in Kazakhstan, Kirgizstan and Tajikistan to develop national robust and adequate 

regulations and procedures, taking into account the international guidance and Norwegian experience in regulatory 

support projects in Russia. Specific expected results in the project period include: a threat assessment report identifying 

priority areas for regulatory development, based on the status of current regulatory documents and the hazard presented 

by the different sites and facilities; development of national radioactive waste management strategies in each country; 

development of an enhanced regulatory framework. 

7) 40259 – Strengthening the Nuclear Security System of Central Radioactive Waste Processing and Storage 

Facility at AERE 

Abdus Sattar Mollah, Bangladesh Atomic Energy Commission (Bangladesh) 

In the aftermath of the events surrounding September 11, 2001, international organizations have called upon 

worldwide users and holders of radioactive material to increase security measures in order to prevent a loss of control 

through acts of theft or sabotage. <strong>The</strong> universal concern is that unsecured material could end up in the possession of a 

terrorist group for the purpose of constructing and detonating a radiological dispersion device or “dirty bomb”. 

Regardless of the security or risk classification criteria used, source material used in all categories of radioactive 

materials is categorized as requiring the highest level of protection to prevent unauthorized access and acquisition. In 

2006, the BAEC formalized specific protection requirements for all licensees of high strength radioactive materials 

within their jurisdiction. <strong>The</strong> basic requirements include establishing physical security perimeters, controlling and 

monitoring access through those perimeters, and verifying that persons authorized access to the source material are 

trustworthy and reliable. <strong>The</strong> practical implementation of these requirements necessitates the involvement of a security 

professional to provide guidance in the creation and implementation of a comprehensive facility security plan. <strong>The</strong> 

following activities have been done to controlling access to the central radioactive processing and storage: 

• Control access into the facility building, 

• designating control entrance points for employees, visitors and contractors, 

• equipping all perimeter egress-only doors with alarms, 

• requiring all dock doors to be closed or openings secured when not in use, 

• Additional access control to the personnel entry door, 

• Control access to the source material. 

For the most part this is achieved through shield design and required safety interlock systems and alarms. <strong>The</strong> shield 

structure is substantial in both size and thickness, while entry or exit through the product portal is generally difficult. This 

paper presents approaches that may be used to improve nuclear security system at Central Radioactive Waste Processing 

and Storage Facility including increased vigilance, education and awareness, and physical changes to the facility. 

SESSION H1: : National and <strong>International</strong> Programs (1) 

1) 40097 – Overview of NUMO's policy for implementing safe geological disposal and developing supporting 

technologies 

Hiroyuki Tsuchi, Kenichi Kaku, Katsuhiko Ishiguro, Akira Deguchi, Yoshiaki Takahashi, NUMO (Japan) 

Based on the Act on Final Disposal of Specified Radioactive Waste (Final Disposal Act), the Nuclear Waste 

Management Organization of Japan (NUMO) was established in October 2000 as the implementing organization for 

geological disposal of high-level radioactive waste. In December 2002, NUMO issued a nationwide call for volunteers to 

initiate the siting process. With the revision of the Final Disposal Act in 2007, some types of TRU waste were included 

as waste destined for geological disposal and thus fall within the remit of NUMO. 

NUMO has been developing the technologies required for implementation of the project and has carried out a range of siting and 

performance assessment (PA) activities. However, no application has been received as yet from a volunteer municipality. Together 

with the national government, NUMO is now making increased efforts to obtain public acceptance for initiating literature surveys (the 

first step in the siting process). In this connection, NUMO decided to prepare a 2010 technical report as a tool for improving the 

understanding of the geological disposal project by the stakeholders. <strong>The</strong> report will present the safety concept that describes how 

NUMO aims to achieve safe geological disposal through ca.100-year project and will document the progress made with developing the 

technologies that support the safety concept. 

Three policies for ensuring safety are followed by NUMO, namely a) staged project implementation and decision making based 

on iterative confirmation of safety, b) project implementation based on reliable technologies (or Best Available Technologies - BAT) 

and c) technical activities for building confidence in NUMO’s safety concept. One of the highlights of the report is the introduction of 

a “roadmap” which includes all project milestones, goals in each stage and main activities for achieving these goals. <strong>The</strong> roadmap is 

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also being used to identify the technologies required for implementing the geological disposal project, thus providing a guideline for 

evaluating currently available technologies and identifying needs for further technology development. <strong>The</strong> technologies developed to 

date and the plan for further technical development are explained in this paper. 

2) 40150 – Stepwise Site Selection in Switzerland - Sectoral Plan: Status and Outlook 

Thomas Ernst, Markus Fritschi, Stratis Vomvoris, Nagra (Switzerland) 

<strong>The</strong> starting point is the nomination of siting regions that fulfil the criteria for long-term safety and engineering 

feasibility defined in the Sectoral Plan. In October 2008, Nagra proposed a total of 6 siting regions for the LLW 

repository and three for the HLW repository; the latter would also be suitable for shared use with the surface facilities 

and part of the access tunnels for the two types of repositories. 

<strong>The</strong> review of Nagra’s proposals by the safety authority (ENSI – the Swiss Federal Nuclear Safety Inspectorate) and 

its supporting commission was completed in February 2010. <strong>The</strong> decision-making process foresees evaluation of the 

review and recommendations of ENSI by the various agencies at the governmental, cantonal and local level, an open 

public consultation and finally, a resolution of the comments received by the Swiss Federal Office of Energy, a 

recommendation to the Federal Government and a decision by the Federal Government. 

At the same time, the procedures and guidelines for the selection of the (at least) two potential sites for each of the 

repositories are being finalised by the safety authorities. Although the safety and engineering criteria remain the same, 

Stage 2 foresees a provisional safety analysis for each potential site as one of the criteria to be used in the site selection 

process. 

This paper will highlight the main findings of the review by the authorities and Nagra’s response to issues raised. 

<strong>The</strong> next steps and preparatory activities for the initiation of Stage 2 will also be described, as well as how the criteria 

and guidelines specified by ENSI will be applied by Nagra in order to meet the requirements for a successful completion 

of Stage 2. 

3) 40084 – Site Selection for a Geological Disposal in France: an Approach of Convergence 

Gerald� Ouzounian, Roberto Miguez, Jean-Louis Tison, Andra (France) 

On December 1991, the French National Assembly passed the French Waste Management Research Act, 

authorizing a 15 year research program of three options for HLW: separation and/or transmutation, long-term storage, 

and geologic disposal. On June 2006, the “Planning Act on the sustainable management of radioactive materials and 

waste" sets a new framework and new aims to the above mentioned options. 

This paper deals only with the geologic disposal research program. In a step by step approach, this program has 

been broken down into three phases having intermediate goals : site selection for an Underground research Laboratory 

(URL), potential disposal feasibility, potential reversible disposal design. 

<strong>The</strong> first step of the research program aimed at URL site selection. From 1994 to 1996, Andra carried out 

geological-characterization work in four departments. This enabled to make the Request for Licensing and Operation of 

the laboratory facility on three sites. During this phase, wells, 2D seismic campaigns and land studies of geologic 

outcrops were the essential activities. <strong>The</strong> result was the selection of the most suitable site for the implementation of an 

underground laboratory. Main results on Bure URL will be presented in the paper. 

In the second phase the research program targeted the safety and technical feasibility of a potential reversible 

disposal somewhere in Meuse and Haute Marne D departments site, chosen by the government in 1998. Andra conducted 

geologic survey during the URL shaft construction and experiments in drifts at depths of 445 and 490 m. This program 

allowed consolidating the knowledge already acquired: geological environment, stability of the rock, containment 

properties and it confirms that the rock will maintain its qualities. <strong>The</strong> 2005 Progress Report presents the results of this 

phase. <strong>The</strong> main conclusion is that a potential disposal facility may be safely constructed over a zone around the URL, 

called transposition zone (about250 km2). <strong>The</strong> paper will present the most important results in this phase. 

From 2006, the third phase of the program, the activities were oriented, inside the transposition zone, to determine a 

smaller zone in which à potential disposal facility could be designed. In 2009, Andra reported to the French authorities a 

proposal describing such a zone. In this paper the main results of this phase will be presented. Finally, next steps towards 

a final implementation will be given. 

SESSION H2: Transportation, Storage and Waste Treatment 

1) 40155 – Support of the Nuclear Research Institute Rez plc of the Shipment of Spent Nuclear Fuel from 

Research Reactors to the Russian Federation for Reprocessing in the Frame of the RRRFR Program 

Josef Podlaha, Karel Svoboda, Nuclear research institute plc. (Czech Republic) 

In 2007, spent nuclear fuel (SNF) from the Nuclear Research Institute Rez plc (NRI) was shipped to the Russian 

Federation for reprocessing. A large amount of SNF of Russian origin has been accumulated after 50 years of research 

reactor operation. 

<strong>The</strong> shipment was realized in the frame of the Russian Research Reactor Fuel Return (RRRFR) program under the 

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US-Russian Global Threat Reduction Initiative (GTRI). SNF shipment from NRI to the Russian Federation (RF) 

represented a very complex and complicated technical, legal and contractual scope of work. 

<strong>The</strong> SNF shipment has been realized under specific conditions: 1. High capacity SKODA VPVR/M casks were used 

for transportation for the first time. 2. For the first time, high enriched uranium SNF from a research reactor has been sent 

to the RF from a European Union country under the appropriate intergovernmental agreements, legal regulations and 

conditions. 

NRI also participates in shipments of SNF from other countries within the framework of the RRRFR program. 

NRI`s participation consists of: - SKODA VPVR/M casks leasing, including service and maintenance inspections of the 

casks, - transportation of the casks to the research reactor site, -providing cask documentation in support of development 

of licenses, certificates, etc. for authorization of the shipment, - training of personnel in cask use and SNF loading, - 

technical oversight and expertise during the cask handling, fuel loading and cask closing and sealing, - the drying and 

helium leak testing of casks, -the return transportation of the empty casks from the Russian/Ukrainian border to NRI. 

NRI participated in shipments of SNF from Bulgaria and Hungary in 2008, from Poland in 2009 and 2010 adn from 

Ukraine in 2010. Shipments from Belarus and Serbia are planned in 2010. <strong>The</strong> second shipment of the residue of high 

enriched SNF from NRI after changeover of the reactor operation to low enriched fuel will be implemented in 2013. 

<strong>The</strong> experiences gained during the SNF transportation are described in the paper together with the present and 

future NRI activities in support of the SNF shipment from other countries. 

2) 40177 – Long Term Storage of Nuclear Spent Fuel as Key Role of Japan's Nuclear Fuel Cycle until 2100: 

Cost and Benefit 

Tadahiro Katsuta, Meiji University (Japan) 

Political and technical advantages to introduce spent nuclear fuel interim storage into Japan's nuclear fuel cycle are 

examined. Once Rokkasho reprocessing plant starts to be operated, 80,000 tHM of spent Low Enriched Uranium (LEU) 

fuel must be stored in an Away From Reactor (AFR) interim storage site until 2100. If a following reprocessing plant 

starts the operation, the Spent LEU reaches its peak of 30,000 tHM before 2050, and then decreases until the end of the 

following reprocessing plant operation. Throughput of the following reprocessing plant is assumed as twice of Rokassho 

reprocessing plant case, namely 1,600tHM/year. On the other hand, three times of final disposal site of High Level 

Nuclear Waste (HLW) will be necessary with this condition. Besides, large amount of plutonium surplus will occur, even 

if First Breeder Reactors (FBR)s consume the plutonium. <strong>The</strong> maximum of plutonium surplus will reach almost 500 tons 

of plutonium. <strong>The</strong>se results indicate that current nuclear policy does not solve the spent fuel problems but even 

complicate them. Thus, reprocessing policy could put off the problems in spent fuel interim storage capacity and other 

fears could appear such as difficulties in large amount of HLW final disposal management or separated plutonium 

management. If there is no reprocessing or MOX use, the amount of spent fuel will reach over 115,000 tones at the year 

of 2100. However, the spent fuel management could be simplified and also the cost and the security would be improved 

by using an interim storage primarily. 

3) 40285 – TRU Recycling Options for increasing Protected Plutonium Production of FBR 

Sidik Permana, Mitsutoshi Suzuki, JAEA (Japan) 

<strong>The</strong> embodied challenges for introducing closed fuel cycle are utilizing advanced fuel reprocessing and fabrication 

facilities as well as nuclear nonproliferation aspect. Optimization target of advanced reactor design should be maintained 

properly to obtain high performance of safety, fuel breeding and reducing some long-lived and high level radioactivity of 

spent fuel by closed fuel cycle options. In this paper, the contribution of loading trans-uranium to the core performance, 

fuel production, and reduction of minor actinide in high level waste (HLW) have been investigated during reactor 

operation of large fast breeder reactor (FBR). Excess reactivity can be reduced by loading some minor actinide in the 

core which affect to the increase of fuel breeding capability, however, some small reduction values of breeding capability 

are obtained when minor actinides are loaded in the blanket regions. As a total composition, MA compositions are 

reduced by increasing operation time. Relatively smaller reduction value was obtained at end of operation by blanket 

regions (9 %) than core regions (15 %). In addition, adopting closed cycle of MA obtains better intrinsic aspect of 

nuclear nonproliferation based on the increase of even mass plutonium in the isotopic plutonium composition. 

4) 40132 – Transuranic (TRU) Waste Volume Reduction Operation at a Plutonium Facility 

Michael Cournoyer, Archie E. Nixon, Keith W. Fife, Arnold M. Sandoval, 

Vincent E. Garcia, Robert L. Dodge, LANL (USA) 

Programmatic operations at the Los Alamos National Laboratory Plutonium Facility (TA-55) involve working with 

various amounts of plutonium and other highly toxic, alpha-emitting materials. <strong>The</strong> spread of radiological contamination 

on surfaces, airborne contamination, and excursions of contaminants into the operator’s breathing zone are prevented 

through use of a variety of gloveboxes (the glovebox, coupled with an adequate negative pressure gradient, provides 

primary confinement). Size-reduction operations on glovebox equipment are a common activity when a process has been 

discontinued and the room is being modified to support a new customer. <strong>The</strong> Actinide Processing Group at TA-55 uses 

one-meter or longer glass columns to process plutonium. Disposal of used columns is a challenge, since they must be 

43

Abstracts 

size-reduced to get them out of the glovebox. <strong>The</strong> task is a high-risk operation because the glass shards that are generated 

can puncture the bag-out bags, leather protectors, glovebox gloves, and the worker’s skin when completing the task. One 

of the Lessons Learned from these operations is that Laboratory management should critically evaluate each hazard and 

provide more effective measures to prevent personnel injury. A bag made of puncture-resistant material was one of these 

enhanced controls. We have investigated the effectiveness of these bags and have found that they safely and effectively 

permit glass objects to be reduced to small pieces with a plastic or rubber mallet; the waste can then be easily poured into 

a container for removal from the glovebox as non-compactable transuranic (TRU) waste. This size-reduction operation 

reduces solid TRU waste volume generation by almost 2½ times. Replacing one-time-use bag-out bags with multiple-use 

glass crushing bags also contributes to reducing generated waste. In addition, significant costs from contamination, 

cleanup, and preparation of incident documentation are avoided. This effort contributes to the Los Alamos National 

Laboratory Continuous Improvement Program by improving the efficiency, cost-effectiveness, and formality of glovebox 

operations. In this report, the technical issues, associated with implementing this process improvement are addressed, the 

results discussed, effectiveness of Lessons Learned evaluated, and waste savings presented. 

5) 40247 – Disposition of Transuranic Residues from Plutonium Isentropic Compression Experiment (Pu-ICE). 

LA-UR 10-04699. 

Kapil Goyal, David M. French, LANL (USA); Betty J. Humphrey, Weston Solutions, Inc. (USA); 

Jeffry Gluth, Sandia National Laboratories/NM (USA) 

In 1992, the U.S. Congress passed legislation to discontinue above- and below-ground testing of nuclear weapons. 

Because of this, the U.S. Department of Energy (DOE) must rely on laboratory experiments and computer-based 

calculations to verify the reliability of the nation’s nuclear stockpile. <strong>The</strong> Sandia National Laboratories/New Mexico 

(SNL/NM) Z machine was developed by the DOE to support its science-based approach to stockpile stewardship. 

SNL/NM researchers also use the Z machine to test radiation effects on various materials in experiments designed 

to mimic nuclear explosions. Numerous components, parts, and materials have been tested. <strong>The</strong>se experiments use a 

variety of radionuclides; however, plutonium (Pu) isotopes with greater than ninety-eight percent enrichment are the 

primary radionuclides used in the experiments designed for stockpile stewardship. In May 2006, SNL/NM received 

authority that the Z Machine Isentropic Compression Experiments could commence. 

Los Alamos National Laboratory (LANL) provided the plutonium targets and loaded the target assemblies, which 

were fabricated by SNL/NM. LANL shipped the loaded assemblies to SNL/NM for Z machine experiments. Three 

experiments were conducted from May through July 2006. <strong>The</strong> residues from each experiment, which weighed up to 913 

pounds, were metallic and packaged into a respective 55-gallon drum each. Based on a memorandum of understanding 

between the two laboratories, LANL provides the plutonium samples and the respective radio-isotopic information. 

SNL/NM conducts the experiments and provides temporary storage for the drums until shipment to LANL for final waste 

certification for disposal at the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. This paper presents a 

comprehensive approach for documenting generator knowledge for characterization of waste in cooperation with 

scientists at the two laboratories and addresses a variety of topics such as material control and accountability, safeguards 

of material, termination of safeguards for eventual shipment from SNL/NM to LANL, associated approvals from 

DOE-Carlsbad Field Office, which governs WIPP and various notifications. It portrays a comprehensive approach 

needed for successful completion of a complex project between two national laboratories. 

6) 40188 – A Milestone in Vitrification - the Replacement of a Hot Metallic Crucible with a Cold Crucible 

Melter in a Hot Cell at the La Hague Plant 

Sphiee Robert, Benoit Carpentier,SGN (France); Florence Gassot Guilbert, SGN Service procédé (France); 

Sandrine Naline, AREVA (France); Frédéric Gouyaud, AREVA NC (France); Christophe Girold, CEA (France) 

Vitrification of high-level waste is the internationally recognized way to optimize the conditioned High Level Waste 

to be disposed of. 

AREVA’s successful operation of AVM at the Marcoule plant, R7 and T7 at the La Hague Plants demonstrates the 

capabilities and experience of the Group to deploy innovative high level waste (HLW) processing technologies in 

industrial facilities, in partnership with R&D (CEA) and AREVA engineering (SGN). CEA and AREVA engineering 

team have continuously improved the hot metallic crucible melter vitrification technology through operational feedback 

aswell as ongoing research and developement. <strong>The</strong>y have led the way in the development and implementation of Cold 

Crucible Induction Melter technology (CCIM). 

<strong>The</strong> cold crucible is a compact water-cooled melter in which the radioactive waste and the glass are molten by direct 

high frequency induction. This technology can handle with highly corrosive solutions and high operating temperature, 

which offers, among others advantages: o a great flexibility in matrix compositions, o the increase of the industrial glass 

production capacity, ° increase of melting temperature, ° allow higher waste loading, ° extend to a wide and varied waste 

types 

In order to take advantage of CCIM, a new project, called “Vitrification 2010”, was launched in 2005 with the 

objective to put in commercial operation a CCIM in one of the existing and active R7 vitrification line in 2010. <strong>The</strong> main 

challenge for the project team is to replace the hot metallic crucible melter with the CCIM: o in an existing hot cell which 

is in operation for 20 years with high level waste solutions, o with respect to the current hot cell’s layout (calciner, 

through-wall for utility for example…). 

44

Abstracts 

<strong>The</strong> paper will illustrate why this event is a major milestone in vitrification and how the “Vitrification 2010” project 

succeeds in record time and without any impact on the La Hague Plant production. 

7) 40265 – Adaptation of CCIM technology for HLW treatment. Results of research and development 

Vladimir Lebedev, Alexander Kobelev, Fyodor Lifanov, Sergey Dmitriev, SIA Radon (Russia) 

Results of research of possibility of Cold Crucible Inductive Melter (CCIM) technology application for HLW 

treatment (on examples of Savannah River (USA) HLW and PA “Mayak” (Russia) HLW), carried out in Moscow SIA 

RADON, and also results of development of new perspective facility for appliance in lab scale and then for real HLW 

treatment, are shown in this report. From 2003 till 2010 the research of possibility of different HLW treatment with 

CCIM technology are carried out. <strong>The</strong> principal possibility of vitrification of wastes with different compositions in the 

cold crucible was shown. <strong>The</strong> features of appliance of various materials, natural and industrially produced, as glass 

forming and modifying additions were explored. Charge compositions, made with using each researched sample of HLW 

imitator, were formulated and maximal HLW containing in charge was determined. <strong>The</strong>n the samples of glass obtained 

were explored. <strong>The</strong> main technological features of melting process were determined and principals of organization and 

control of the process were formulated. <strong>The</strong> pattern of cold crucible for HLW imitators treatment being included in 

experimental stand in PA “Mayak” was created. Also the System of Automatic Control for crucible was created. <strong>The</strong> 

tests and research of equipment work were carried out. On the base of analysis of data gained earlier and newly obtained 

the main requirements to designing of inductive melters and additional equipment, intended for real HLW treatment, 

were worked out. <strong>The</strong>se requirements seem like a part of Task for designing of experimental-industrial HLW treatment 

plant, that is contemplated to built in PA “Mayak” in nearest time. 

SESSION H3: Panel "Radwaste Human Resource Development to Support the Nuclear Renaissance" 

Abstract Not Required 

SESSION D1 : National and <strong>International</strong> D&D Programs 

1) 40003 – Westinghouse PWR and BWR Reactor Vessel Segmentation Experience in Using Mechanical 

Cutting Process 

Joseph Boucau, Stefan Fallström, Per Segerud, Paul Kreitman, Westinghouse Electric Belgium (Belgium) 

Some commercial nuclear power plants have been permanently shut down to date and decommissioned using 

dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of 

reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven 

methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based 

on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal 

disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 

for both PWR’s and BWR’s and its process has been continuously improved over the years. This paper will describe the 

sequential steps required to segment, separate, and package each individual component, based on this mechanical cutting 

method. Detailed planning is essential to a successful project, and typically a “Segmentation and Packaging Plan” is 

prepared to document the effort. <strong>The</strong> usual method is to start at the end of the process, by evaluating the waste disposal 

requirements imposed by the waste disposal agency, what type and size of containers are available for the different 

disposal options, and working backwards to select the best cutting tools and finally the cut geometry required. <strong>The</strong>se 

plans are made utilizing advanced 3-D CAD software to model the process. Another area where the modeling has proven 

invaluable is in determining the logistics of component placement and movement in the reactor cavity, which is typically 

very congested when all the internals are out of the reactor vessel in various stages of segmentation. <strong>The</strong> main objective 

of the segmentation and packaging plan is to determine the strategy for separating the highly activated components from 

the less activated material, so that they can be disposed of in the most cost effective manner. Usually, highly activated 

components cannot be shipped off-site, so they must be packaged such that they can be dry stored with the spent fuel in 

an Independent Spent Fuel Storage Installation (ISFSI). Less activated components can be shipped to an off-site disposal 

site depending on space availability. Several of the plants dismantled to date in the US have repackaged the less activated 

waste back into the reactor vessel and shipped the entire assembly to the disposal site. Decisions like these can be driven 

by many factors such as disposal costs, transportation logistics, licensing fees, etc., but will have a significant impact on 

the segmentation and packaging plan so they must be considered early in the planning phase. All segmentation tools are 

remotely controlled since the mechanical segmentation projects that Westinghouse has executed, so far, have been 

performed under water due to the high radiation levels. ALARA and personal safety is the number one priority during the 

site work. <strong>The</strong> complexity of the work requires well designed and reliable tools. Westinghouse has optimized the 

technologies from its experiences accumulated over the years. Its main focus has always been to improve tool handling 

and cutting speed, water cleanliness, fail-safe and safety aspects. Different band saws, disc saws, tube cutters and 

45

Abstracts 

shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is 

very suitable for submerged applications. Mechanical cutting has a number of advantages compared to other cutting 

techniques. ? <strong>The</strong> technique produces almost no secondary waste. ? <strong>The</strong> visibility during cutting is very good because the 

cutting produces only a negligible amount of micro particles. ? Chips from the cutting process falls down to the bottom of 

the cutting pool and are easy to collect. ? No gases are produced that can cause airborne contamination. ? <strong>The</strong> technique 

is safe and reliable. ? All reactor internal sizes, materials and thicknesses can be cut. 

Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation 

of internals. Westinghouse continues to develop new methods and products in order to further reduce the waste volume. 

In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation 

process, based on actual experience from the work that has been completed to date. 

2) 40130 – EPRI Nuclear Power Plant Decommissioning Technology Program 

Karen Kim, Sean Bushart, Mike Naughton, Richard McGrath, Electrict Power Research Institute (USA) 

<strong>The</strong> Electric Power Research Institute (EPRI) is a non-profit research organization that supports the energy industry. 

<strong>The</strong> Nuclear Power Plant Decommissioning Technology Program conducts research and develops technology for the safe 

and efficient decommissioning of nuclear power plants. 

Several USA nuclear power plants shut down and entered active decommissioning in the 1990s. At that time the 

decommissioning of a commercial nuclear power plant was a first of a kind evolution; nuclear power plant 

decommissioning is a combination of industrial decommissioning with the complexities of radioactive materials and 

waste management. <strong>The</strong> first plants to undergo this process encountered various challenges related to decommissioning 

planning and project management, removal of large radioactive components, radiation protection, site remediation, final 

status survey, and license termination. <strong>The</strong> USA regulatory framework for nuclear power plant decommissioning 

developed and evolved along with these first decommissioning projects. As of 2009 these nuclear power plants have 

successfully completed decommissioning activities and have achieved license termination. Two major contributors to the 

successful completion of these projects were a) the development of technology specifically tailored to tasks unique to 

nuclear plant decommissioning and b) learning from each other’s experiences. 

<strong>The</strong> EPRI Nuclear Power Plant Decommissioning Technology Program conducts research and develops technology 

that addresses the range of decommissioning tasks from pre-planning, site characterization, removal of large components, 

site remediation, to license termination. A key feature of the EPRI Program is the collection and transfer of lessons 

learned and experiences from the major decommissioning projects conducted to date. <strong>The</strong> EPRI Program conducts 

research and development based on the needs of its utility members; the EPRI Program membership consists of utilities 

from the USA, France, Spain, Sweden, Italy, Taiwan, and Japan. EPRI is the mechanism through which these 

international utilities can conduct collaborative research and development and learn from each other’s experiences and 

lessons learned. In addition to collaborative research and development, the EPRI Program Team of decommissioning 

experts provides direct and site-specific support to its member utilities to facilitate knowledge transfer and 

decommissioning project optimization. 

This presentation will provide information about the EPRI Nuclear Power Plant Decommissioning Technology 

Program and discuss USA and <strong>International</strong> Decommissioning experiences to date. 

3) 40253 – Tokai-1 Decommissioning Project 

Keizaburou Yoshino , JAPC (Japan) 

Tokai-1 is the oldest and historical commercial Magnox reactor in Japan, which had started commercial operation in 

1966. <strong>The</strong> unit had helped introduction and establishment of the construction and operation technologies regarding 

nuclear power plant at early stage in Japan by its construction and operating experiences. However, <strong>The</strong> Japan Atomic 

Power Company (JAPC), the operator and owner of Tokai-1, decided to cease its operation permanently because of a 

fulfillment of its mission and economical reason. <strong>The</strong> unit was finally ceased in March 1998 after about 32 year operation. 

It took about three years for removal of all spent fuels from the site, and then decommissioning started in 2001. JAPC, 

always on the forefront of the nation’s nuclear power generation, is now grappling Japan’s first decommissioning of a 

commercial nuclear power plant, striving to establish effective, advanced decommissioning. <strong>The</strong> decommissioning for 

Tokai-1 was scheduled as 20 years project. At the beginning, the reactor was started to be in a static condition for ten 

years (“safe storage period”). While the reactor had been safely stored, the phased decommissioning works started from 

non-radioactive or low radioactive equipments toward high radioactive equipments. First five years of the project, JAPC 

concentrated to drain and clean spent fuel cartridge cooling pond and to remove conventional equipments such as turbine, 

feed water pump and fuel charge machine as planed and budgeted. 

From 2006, the project came into new phase. JAPC has been trying to remove four Steam Raising Units (SRUs). 

<strong>The</strong> SRUs are huge component (7ton, �6.3m, H24.7m) of the Gas Cooling Reactor (GCR) and inside of the SRUs are 

radioactively contaminated. It is concerned that workers are required safety and minimizing contamination areas during 

SRU removal. <strong>The</strong>refore, JAPC is developing and introducing Jack-down method and remote control multi functional 

removal system. This method is the method by which to remove the SRUs in turn from the bottom by lifting the SRU by 

a large jack and cutting for removing SRU is done remotely with this system. <strong>The</strong> system enables cutting and holding not 

only SRU body but also internals. This technology and experiences would be useful for the reactor removal in the near 

future. 

46

Abstracts 

4) 40289 – Activities of the OECD/NEA WPDD in the Field of Decommissioning 

Claudio Pescatore, Patrick O'Sullivan, OECD/NEA 

<strong>The</strong> OECD/NEA seeks to assist its member countries in developing strategies for the management of all types of 

radioactive material, including waste, that are safe and sustainable and that meet the broad needs of society - with 

particular emphasis on the management of long-lived waste and spent fuel and on decommissioning of disused nuclear 

facilities. <strong>The</strong> programme of work in the area of decommissioning is supervised by the Radioactive Waste Management 

Committee (RWMC) and carried out by the Working Party on Decommissioning and Dismantling (WPDD). <strong>The</strong> latter is 

made of senior representatives from regulatory authorities, decommissioning organisations, policy making bodies, and 

researchand-development institutions from the NEA countries. It includes also representatives of the IAEA and of the the 

European Commission. <strong>The</strong> WPDD supports the RWMC by keeping under review the policy, strategic, and regulatory 

aspects of decommissioning of phased-out nuclear installations. Its scope of work includes decommissioning and 

dismantling of shutdown facilities up to and including the release of the site, but excluding fuel removal, removal of 

nuclear processing fluids, post-operational clean out of fuel residues and removal of operational wastes. <strong>The</strong> work 

programme comprises activities in the following key areas: Policy, regulation and strategy; Funding and costs; 

Techniques; Decommissioning materials management and site release; and Human and organisational factors. Recent 

trends in the above areas, as noted in a recent analysis by the WPDD, include: • Policy, regulations and strategy – most 

countries have adopted policies for decommissioning and for funding provisioning but a range of options were being 

applied to the decommissioning of nuclear power plants (early dismantling, safe store and entombment), • Funding and 

costs – cost calculations for decommissioning are very sensitive to the assumed end-state and to the levels of 

contaminations on the sites and may be greatly influenced by stakeholder requirements, • Techniques – the focus of 

current R&D is on the development of innovative technology for segmentation, dismantling and concrete 

decontamination, together with better instrumentation for material management and control • Materials management and 

site release – there is a significant acceleration in the number of decommissioning projects, including provision of 

infrastructure for storage of materials and for undertaking clearance and recycling of disused materials; • Human and 

organisational factors – efforts are being made to structure decommissioning contracts to optimise the supply chain 

relationships, with greater transparency and better communication amongst all concerned parties; and • Stakeholder 

participation and knowledge management – there is a range of approaches to stakeholder involvement in developing 

decommissioning plans and to rule making. In general, this issue is gaining momentum. A number of specific activities, 

projects and reports address the above areas. <strong>The</strong>y will be reviewed for the ICEM-2010 audience 

5) 40307 – French Decommissioning Feedback Experience and Lessons Learned 

Jean-Guy Nokhamzon, CEA (France); Patrick O'Sullivan, OECD/NEA 

<strong>The</strong> French (CEA, EDF and AREVA) experience in decommissioning of nuclear facilities goes back many decades 

and relates to many installations of very different types. 

Some twenty facilities were dealt with by 2009, corresponding to around half of all the nuclear facilities 

permanently closed up to date, beginning in March 2001 with the decommissioning of AT1 facility at La Hague, the pilot 

plant used by the CEA in the seventies for the reprocessing of spent fuel from fast neutron reactors (former IAEA stage 3, 

excluding the civil engineering structures), as well as the blasting of G1 stack at Marcoule, on 19 July 2003 and ending 

with the total demolition of the research reactors Triton and Néréide in Fontenay aux Roses in 2004 and the final steps 

for the total decommissioning of Siloë, Siloette and Mélusine in Grenoble research Centre by 2012. 

All the work done demonstrates the following: 

1) Decommissioning can and has been done in a safe, cost-effective and environmentally friendly manner. 

2) Current technologies have demonstrated their effectiveness and robust performance in numerous 

decommissioning activities. Feedback experience on design, construction and operation is a considerable help for 

reliable planning, cost evaluation and successful realization of a decommissioning project.� 

3) During decommissioning radiological risks are very small in comparison to non-radiological risks. 

4) <strong>The</strong> dissemination of best practices and sharing of information in international workshops, conferences and 

especially within working groups as the OECD/NEA/CPD and IAEA/IDN has proven to be a good basis for an 

effective cooperation and support to master new challenges on decommissioning projects.� 

5) Future challenges will require further international cooperation to establish sustainable regulations and 

guidance to achieve objectives without being burdensome or overly conservative. 

6) A consistent, internationally accepted rationale is necessary for the elaboration of concepts and for the 

derivation of numerical values on clearance, exemption and authorized releases. 

7) With decommissioning moving towards being a fully mature industrial process, increased dialogue among 

regulators, implementers and international standards organisations is necessary. 

6) 40032 – Decommissioning Planning for Swedish Operating NPPs� 

Gunnar Hedin, Mathias Edelborg, Niklas Bergh, Westinghouse Electric Sweden (Sweden); Jan Carlsson, Fredrik 

de la Gardie, SKB (Sweden) 

47

Abstracts 

Decommissioning studies have been carried out for the three BWR units of Oskarshamn and the three BWR units of 

Forsmark nuclear power plants. <strong>The</strong> final closure of these units is far ahead but anyhow there has been a need for 

developing a more general decommissioning planning basis. <strong>The</strong> main objectives of the studies have been to establish an 

estimate of the waste amounts arising from these units during decommissioning and dismantling as well as providing a 

firm basis for funding of the decommissioning phase for these units. <strong>The</strong> waste amounts will be used when designing a 

separate repository for decommissioning waste of the same type as the existing facility for final disposal of short-lived 

low- and intermediate level waste, the SFR, at Forsmark, Sweden. <strong>The</strong> broader studies are also used for verification that 

the existing national decommissioning fund is of an adequate size. 

<strong>The</strong> studies have been performed by Westinghouse in cooperation with the utilities of Forsmark (FKA) and 

Oskarshamn (OKG), on behalf of SKB, the utility-owned Swedish waste management organization, responsible for 

coordination of the national waste fund as well as for designing, building and operation of waste management facilities. 

Each of the studies contains a general description of the plant with the purpose to characterize them to facilitate 

decommissioning project comparative studies of different plants. Also, a detailed inventory of all materials in the plant 

have been put together and modeling of radioactivity data have been carried out to establish typical contamination levels 

for the main parts of the plant. <strong>The</strong> procedures and technologies foreseen to be used for the future dismantling and waste 

management work are briefly described based on standard, presently easily available techniques. 

<strong>The</strong> combination of the materials inventory data and the radiological data is used to calculate the number of 

different waste containers that will be produced during dismantling of the plant. Standard 20 feet ISO containers will be 

used for the low-level waste while special steel containers that fit the transport overpack system will be used for 

intermediate level waste. <strong>The</strong> long-lived waste will be loaded into thick-walled steel containers. 

<strong>The</strong> inventory data will also be used by experience models that calculate work hours for taking care of all the 

different types of plant components. <strong>The</strong> working time estimates are then combined, together with general duration data 

for different activities during plant decommissioning, into a time schedule for the complete program, from initial 

planning and preparatory activities to non-radioactive building demolition and site restoration. 

Costs are then estimated for each of the work breakdown structure (WBS) elements of the time schedule. 

Summarized it provides a budgetary estimate for the complete decommissioning program. By rearranging the elements of 

the cost estimate according to the IAEA/EC/OECD-NEA common matrix (the “Yellow Book”), cost comparison with 

other plants is facilitated. 

<strong>The</strong> paper will provide detailed data of waste characterization and amounts, as well as of time schedules and cost 

estimates for the main part of the Swedish BWR fleet and show how decommissioning will fit with the existing 

well-developed waste management system of the country. 

7) 40104 – Chernobyl nuclear power plant Decommissioning Program 

Denys Tkachov, Dmytro Stelmakh, Viktor Kuchinskyy, 

State Special Enterprise Chernobyl Nuclear Power Plant (Ukraine) 

After termination of electricity generation in December 2000, Chernobyl nuclear power plant (Ukraine) entered 

final shutdown stage that will be followed by decommissioning of the NPP units. Decommissioning activities will be 

performed on basis of the Decommissioning Program, developed at Chernobyl NPP in 2007. This document includes 

detailed description and substantiation of the selected decommissioning strategy, established sequence, duration and 

basic content of each decommissioning stage, as well as final state of the facility at the end of each stage. An information 

required for using this document as guidance for decommissioning of Chernobyl NPP units is also comprised. <strong>The</strong> 

Decommissioning Program of the Chernobyl NPP units is developed on basis of selected decommissioning strategy, 

namely 

– «deferred dismantling» (SAFESTOR), which includes follow stages: 1) Final shutdown and preservation (10-12 

years); 2) Safe enclosure (up to 30 years); and 3) Dismantling (20 years). Chernobyl nuclear power plant is located in 

area, contaminated by radioactive isotopes during the Accident at Unit 4 in 1986. In addition Unit 3 has common 

structures with Unit 4 (currently the “Shelter Object”). <strong>The</strong>se conditions lead to the classification of the final state of the 

Chernobyl NPP site as “brown spot”. Major criteria of given state are limited release of materials and 

cleanup/decontamination of the structures to the level of limited release from the regulatory control. This article deals 

with issues on implementation of decommissioning activities at Chernobyl NPP units, describe each decommissioning 

stage, its final state and main activities on each state. In addition issues on the management of radioactive waste and 

spent nuclear fuel are included as well as basic Chernobyl NPP units decommissioning cost estimation. 

8) 40273 – Considerations for Grout Formulations in Facility Closures using In Situ Strategies 

John Gladen, Mike Serrato, Chris Langton, Savannah River National Laboratory (USA); 

Andy Szilagyi, US DOE (USA)� 

<strong>The</strong> U.S. Department of Energy is conducting In Situ closures (entombment) at a large number of facilities 

throughout the complex. Among the largest closure actions currently underway is the closure of the P and R Reactors at 

the Savannah River Site. Below grade open spaces in these facilities are being stabilized with grout to ensure the long 

term structural integrity of the facility which ensures the permanent immobilization of residual contamination. 

<strong>The</strong> large size and structural complexity of these facilities present a wide variety of challenges for the identification 

and selection of appropriate fill materials. Considerations for grout formulations must account for flowability, long term 

48

Abstracts 

stability, set times, heat generation and interactions with materials within the structure. <strong>The</strong> large size and configuration 

of the facility necessitates that grout must be pumped from the exterior to the spaces to be filled, which requires that the 

material must retain a high degree of flowability to move through piping without clogging while achieving the required 

leveling properties at the pour site. Set times and curing properties must be controlled to meet operations schedules, 

while not generating sufficient heat to compromise the properties of the fill material. 

<strong>The</strong> properties of residual materials can result in additional requirements for grout formulations. If significant 

quantities of aluminum are present in the facility, common formulations of highly alkaline grouts may not be appropriate 

because of the potential for hydrogen generation with the resultant risks. <strong>The</strong> SRS is developing specialized inorganic 

grout formulations that are designed to address this potential problem. One circum-neutral chemical grout formulation 

identified for initial consideration did not possess the proper chemical characteristics having exceptionally short set times 

and high heat of hydration. Research efforts are directed toward developing formulations that can meet operational 

requirements for chemical compatibility, extended set times and reduced heat generation. 

SESSION R1: Environmental Impacts 

1) 40298 – Main Results of A Remediation of Uranium- and CHC-Contaminated Groundwater 

Jörg Wörner, Sonja Margraf, Walter Hackel, RD-Hanau (Germany) 

<strong>The</strong> ground-water of a former research- and production site for various nuclear fuel-elements has been remediated 

for the last 8 years taking place after an intensive soilremediation by large-scale drilling and additional soil-exchange 

within the saturated zone. <strong>The</strong> remediation-activities were improved successfully by several measures of which one was 

the increase of the plant’s throughput and another the setting of additional remediation-wells at selected places. This 

procedure is mainly based on the regular observation of the remaining contaminants’ concentrations in the ground-water 

and the aquifer’s specific lithography. <strong>The</strong> cleaning process was adopted for a second installation which was taken in 

operation last November. Both installations’ experience will be presented by which about 450.000 m³ ground-water will 

have been extracted by midyear. <strong>The</strong> area affected by the contaminants has been downsized by the remediation which is 

manifested by the results of the monthly monitoring. <strong>The</strong> conventionally set target-values have not yet been reached, but 

the progress will offer the chance to release the remediation out of the rules of the German Atomic Law by the year 2013. 

Details in the ground-water’s Uranium-nuclide composition allow the attribution of the ground-water to two main plumes 

associated to different former production-activities. Blending of the two plumes with respect to their specific 

Uranium-nuclide composition caused by the chosen pump-regime has been observed and is meanwhile a helpful means 

for reaching the set aims on a planned time-scale. Figures for the totally extracted amounts of Uranium and CHC 

(chlorinated hydrocarbons) will be presented which are increasing actually for both contaminants as result of setting 

additional remediation-wells close to the still- observable contamination-centres. Completely remediated areas are 

observed on the other hand where intensive pumping was executed from the very beginning of the remediation-activities. 

<strong>The</strong>se observations give data to estimate the residual amounts of ground-water which have to be treated in the future. <strong>The</strong> 

decrease of the Uranium-contamination is observed as a function of the extracted amount of ground-water whereas the 

CHC-contamination has decreased irregularly due to the fact of its dependence of the pump-regime and the 

remediation-wells’ different distances to the former handling locations. <strong>The</strong> CHC-compound-spectrum shows natural 

attenuation effects in the close down-stream area which can be helpful after achievement of the remediation’s 

target-value for the Uranium and possible finalizing of a continuous pumping procedure. As detailed results were 

published in two former ICEM-papers [1,2] this paper will highlight main results and discuss taken measures to finalize 

the remediation-activities within the next 3 to 4 years. [1] ICEM07-7270 Remediation of a Uranium-contamination in 

groundwater, [2] ICEM09-16244 Remediation of Uranium- and CHC-contaminated groundwater on a former nuclear 

fuel-element production site. 

2) 40267 – Biogeochemical Gradients, Waste Site Evolution, and Implications for Sustained Metal and 

Radionuclide Attenuation in Complex Subsurface Environments 

Karen Skulbal, Justin Marble, Kurt D. Gerdes, US DOE (USA); 

Miles Denham, Karen Vangelas, Savannah River National Laboratory (USA) 

<strong>The</strong> U.S. Department of Energy (DOE) Office of Technology Innovation and Development sponsors applied 

research to improve the understanding of metal and radionuclide behavior in soil, sediment, and groundwater. 

DOE-supported researchers are developing advanced site characterization and simulation capabilities, as well as novel 

remediation tools and technologies to control contaminants over long time frames in highly complex subsurface 

environments. Natural and enhanced attenuation strategies are being studied in the context of subsurface biogeochemical 

gradients, such as variations in pH and redox potential. <strong>The</strong>se gradients are spatially and temporally dynamic and can 

strongly influence metal and radionuclide migration, speciation, and reactivity. Knowledge of gradient evolution is 

essential for site evaluation, initial remedy selection, and technology transitioning to sustainable, low-energy remediation 

approaches such as monitored natural attenuation. Research activities in this area will be discussed using case studies 

from DOE sites. One case study focuses on a former disposal area at the Savannah River Site (SRS) in Aiken, South 

Carolina. Unlined seepage basins at the site received approximately seven billion liters of acidic, aqueous, low-level 

49

Abstracts 

radioactive waste over more than three decades. <strong>The</strong> resulting groundwater plume contains multiple contaminants, 

including nitrate, iodine-129, strontium-90, technetium-99, tritium, and uranium isotopes. Since 1991, groundwater near 

the source area has been analyzed for evidence that plume acidity is naturally attenuating. Other biogeochemical 

gradients are also being evaluated to determine natural waste site evolution and impacts from active remediation systems. 

<strong>The</strong>se systems include a neutralization barrier to mitigate plume acidity and a biostimulation zone to enhance in situ 

reductive metal precipitation. <strong>The</strong> influence of reactive facies on plume dynamics is also being examined using 

geophysical tools to detect these subsurface zones of unique mineralogy, hydrology, and microbiology. Site 

characterization data, plume dynamics, and contaminant behavior are incorporated into an evolving site conceptual model, 

which serves as the foundation for reactive transport modeling. This unique approach for site assessment is expected to 

improve the selection of remedial strategies and decision making for long-term environmental stewardship at SRS and 

other sites. <strong>The</strong> work is complementary to the U.S. Environmental Protection Agency’s recent guidance on monitored 

natural attenuation of inorganics. 

3) 40260 – Current Mercury Distribution and Bioavailability in Floodplain Soils of Lower East Fork Popular 

Creek, Oak Ridge, Tennessee, USA 

Fengxiang X. Han, Yi Su, David L. Monts,� Mississippi State University (USA) 

<strong>The</strong> objectives of this study were to investigate the current status of mercury distribution, speciation and 

bioavailability in the floodplain soils of Lower East Fork Poplar Creek (LEFPC) after decades of US Department of 

Energy’s remediation. Historically as part of its national security mission, the U.S. Department of Energy’s Y-12 

National Security Facility in Oak Ridge, TN, USA acquired a significant fraction of the world’s supply of elemental 

mercury. During the 1950s and 1960s, a large amount of elemental mercury escaped confinement and is still present in 

the watershed surrounding the Y-12 facility. A series of remediation efforts have been deployed in the watersheds around 

the Oak Ridge site during the following years. <strong>The</strong> sampling fields were located in a floodplain of LEFPC of Oak Ridge, 

TN, USA. A series of surface soils (10-20 cm) were sampled from both wooded areas and wetland/grass land. Two 8x8 

m fields were selected in the woodland. Five profiles each consisting of three layers were randomly taken from each field. 

<strong>The</strong> three layers were the surface layer at 0-10cm, subsurface layer at 50-60 cm, and bottom layer at 100-110 cm. Soil in 

both wood and wetland areas was well developed with a clear B horizon. <strong>The</strong> present study clearly shows that the total 

mercury in floodplain soils of LEFPC significantly decreased after the series of remediation. This study confirmed the 

long-term effectiveness of these remediation actions, especially after excavation of highly contaminated floodplain soils. 

However, the average total mercury level of all soil samples collected are in the range of 50-80 mg/kg, still significantly 

above toxic level (> 5mg/kg). Furthermore, contrary to conventional believing, the major mercury form in current soils of 

this particular area of floodplain of LEFPC is mainly in non-cinnabar mercury bound in clay minerals (after decades of 

remediation). <strong>The</strong> floodplains can act both as a medium-term sink and as long-term sources. Native North American 

earthworms (Diplocardia spp.) and adjacent soils were taken from each spot in each field. Our results show strong linear 

relationships between mercury concentrations in earthworms (both mature and immature groups) and non-cinnabar 

mercury form, while cinnabar mercury is less bioavailable to native earthworms. Earthworms may be used as a potential 

mercury ecological bioindicator (bio-marker) for demonstrating mercury bioavailability and ecotoxicity in the ecosystem. 

<strong>The</strong> long-term stability, mobility and bioavailability of mercury contaminants in these floodplains still needs to be 

monitored continuously and closely. 

4) 40262 – Integrated Strategy to Address Hanford's Deep Vadose Zone Remediation Challenges 

Mark B. Triplett, Mark D. Freshley, Michael J. Truex, Dawn Wellman, PNNL (USA); 

Kurt D. Gerdes, Briant L. Charboneau, John G. Morse, Robert W. Lober, US DOE (USA); 

Glen B. Chronister, CH2M Hill Plateau Remediation Company (USA) 

A vast majority of Hanford’s remaining in-ground contaminants reside in the vadose zone of the 200 Area Central 

Plateau, where reprocessing operations occurred. <strong>The</strong> vadose zone at this location is comprised of about 75 meters of 

water-unsaturated, unconsolidated, stratified sediments above groundwater that discharges to the Columbia River. 

Contaminants in this zone originated from intentional discharges to cribs, retention basins, and trenches, and from 

unintended tank waste releases in the tank farms. <strong>The</strong> “deep vadose zone” is defined as the region below the practical 

depth of surface remedy influence (e.g., excavation or surface barrier). At the Hanford Site, this region poses unique 

challenges for characterization and remediation. 

In 2008, the Department of Energy initiated a deep vadose zone treatability test to seek remedies for technetium-99 

and uranium contamination. <strong>The</strong> treatability approach includes laboratory, modeling, and field tests building on previous 

work at the Hanford Site. Initial laboratory and field tests for technetium-99 have focused on use of desiccation which 

could be used in combination with an infiltration barrier to slow the transport of technetium-99 in the subsurface. For 

uranium contamination, reactive gas technologies are being tested as one component of the overall treatability test 

approach. 

More recently, in recognition of the need for a broader strategy, the Department of Energy initiated an integrated 

study of the deep vadose zone to establish a technical basis for addressing risk-driving contaminants and supporting 

selection of remediation approaches for the deep vadose zone that are protective of groundwater. This is a cooperative 

effort that combines the resources and contributions of research scientists, technology developers, and remediation 

contractors. 

50

Abstracts 

<strong>The</strong> objectives of the deep vadose zone strategy are to: 

• Develop sufficient understanding of the nature and extent of deep vadose zone contamination and processes that 

affect fate and transport; 

• Develop predictive capabilities for describing contaminant fate and transport as well as flux from the vadose 

zone to the groundwater; 

• Develop, test and deploy effective methods for remediating contaminated areas; 

• Develop and deploy effective monitoring methods for assessing the performance of remedies and for 

determining the long-term threat of contaminants to the groundwater. 

5) 40235 – Advanced Remedial Methods for Metals and Radionuclides in Deep Vadose Zone Environments 

Dawn Wellman, Shas Mattigod, Ann Miracle, Lirong Zhong, Danielle Jansik, PNNL (USA); 

Susan Hubbard, Yuxin Wu, LBNL (USA); Martin Foote, MSE (USA) 

Deep vadose zone environments can be a primary source and pathway for contaminant migration to groundwater. 

<strong>The</strong>se environments present unique characterization and remediation challenges that necessitate scrutiny and research. 

<strong>The</strong> thickness, depth, and intricacies of the deep vadose zone, combined with a lack of understanding of the key 

subsurface processes (e.g., biogeochemical and hydrologic) affecting contaminant migration, make it difficult to create 

validated conceptual and predictive models of subsurface flow dynamics and contaminant behavior across multiple scales. 

<strong>The</strong>se factors also make it difficult to design and deploy sustainable remedial approaches and monitor long-term 

contaminant behavior after remedial actions. 

Functionally, the methods for addressing contamination must remove and/or reduce transport of contaminants. This 

problem is particularly challenging in the arid western USA where the vadose zone is hundreds of feet thick, rendering 

transitional excavation methods exceedingly costly and ineffective. Delivery of remedial amendments is one of the most 

challenging and critical aspects for all remedy-based approaches. <strong>The</strong> conventional approach for delivery is through 

injection of aqueous remedial solutions. However, heterogeneous deep vadose zone environments present hydrologic and 

geochemical challenges which limit the effectiveness. Because the flow of solution infiltration is dominantly controlled 

by gravity and suction, injected liquid preferentially percolates through highly permeable pathways, by-passing 

low-permeability zones which frequently contain the majority of contamination. Moreover, the wetting front can readily 

mobilize and enhance contaminant transport to the underlying aquifer prior to stabilization. Development of innovative, 

in-situ technologies may be the only way to meet remedial action objectives and long-term stewardship goals. 

Shear-thinning fluids (i.e., surfactants) can be used to lower the liquid surface tension and create stabile foams, 

which readily penetrate low permeability zones. Although surfactant foams have been utilized for subsurface 

mobilization efforts in the oil and gas industry, so far, the concept of using foams as a delivery mechanism for 

transporting remedial amendments into deep vadose zone environments to stabilize metal and long-lived radionuclide 

contaminants has not been explored. Foam flow can be directed by pressure gradients, rather than being dominated by 

gravity; and, foam delivery mechanisms limit the volume of water (< 5% vol.) required for remedy delivery and 

emplacement, thus mitigating contaminant mobilization. We will present the results of a numerical modeling and 

integrated laboratory- / intermediate-scale investigation to simulate, develop, demonstrate, and monitor (i.e. advanced 

geophysical techniques and advanced predictive biomarkers) foam-based delivery of remedial amendments to remediate 

metals and radionuclides in vadose zone environments. 

6) 40152 Water discharge from the Chernobyl NPP cooling pond: the main challenges of the environmental 

protection during post-rehabilitation period 

Dmitri Gudkov, Institute of Hydrobiology (Ukraine) 

One of the problems connected with decommissioning of the Chernobyl NPP is further destiny of the cooling pond 

and planned in 2010 a water discharge from it. Cooling ponds was created on the floodplain part of the Pripyat River 

(main tributary of the Dnieper River) and exhibits an oval in 11-km length and about 2-km width. Pumping of water from 

the Pripyat River provides the constant level of water. In 1986 during Chernobyl NPP accident the huge quantity of 

radionuclides has dropped out directly on a water surface, and also was released with reactor waters and has appeared in 

the cooling pond. Now in the bottom sediments of the cooling pond deposited about (16±3)E+13 Bq of Cs-137, 

(2,4±0,9)E+13 Bq of Sr-90 and (5,3±1,9)E+11 Bq of Pu-239+240. More than 30-50 % of the total radionuclide activity 

deposited in the bottom sediments on the depths less than 7 m. After cessation of pump station during 5-6 years 

(depending on concrete natural conditions) the level of water will decrease on 6-7 m to a level of the Pripyat River and 

on a site of the cooling pond along old river-bed there will be a chain of floodplain lakes It is supposed that water drain 

into the Dnieper River system will not result to any radiological consequences for the environment and human health. 

<strong>The</strong> basic contingent, working in the exclusion zone, will not be subject to influence of carry of radioactive dust from 

surface of the drained sites of the cooling pond, and the contribution of inhalation doze of irradiation will be not essential 

in comparison with impact of other sources of contamination. At the same time for the personnel that will be involved in 

rehabilitation work within the drained area of bottom, the inhalation doze of irradiation can be essential. <strong>The</strong>refore the 

measure on dust depressing will be desirable until the steady obliteration of the open sites of the pond bottom by 

vegetation will take place. At that the phyto-rehabilitation measures, allowing to speed up an obliteration of drained most 

contaminated sites of bottom surface with oozy sediments and being a source of radioactive aerosols eolomotion can be 

51

Abstracts 

effectual. Unfortunately, among expected problems at the water drain in the cooling pond the most complicated one can 

appear not solved problem of safe recycling or natural transformation of higher aquatic plants, phyto- and zooplankton, 

molluscs and fish. <strong>The</strong> planned mode of water drain will result in destruction of huge both on biomass and quantity of 

plant and animal species. Up to 100,000 tons of biota, among which practically all of higher aquatic plants, zoobenthos, 

significant quantity of fish, will be lost. Many species of waterfowls and other birds will lose their habitats, including rare, 

vulnerable and endangered species. Separately it is necessary to mention the problem of dose rate increase on biota due 

to occupancy of drained territories with high density of radioactive contamination, and also in connection with expected 

increase of the radionuclide content in all components of lake ecosystems, which will formed after water drain of the 

cooling pond. 

SESSION M1: Environmental Management 

1) 40086 – Legacy Management: Turning Liabilities into Assets 

Joe Legare, Eric Olson, S.M. Stoller Corporation (USA) 

<strong>The</strong> U.S. Department of Energy (DOE) Office of Legacy Management (LM) is responsible for stewardship of over 

85 sites across the USA dating back to the Manhattan Project and associated with the nuclear weapons production 

mission. This responsibility for the long-term environmental management of these sites includes the key requirements to 

ensure protection of human health and protection of the environment from residual contamination, and to manage over 

100,000 ft3 of associated records. 

<strong>The</strong> Legacy Management program, through its prime contractor, provides program management, field and technical 

support, data collection and analysis and project planning and implementation in support of the DOE to achieve its 

mission. With a defined program baseline of 75 years, and a theoretical baseline that may extend beyond this time period, 

effective programmatic integration of these myriad sites, and appropriate consideration of life-cycle implications to risk 

and cost of stewardship decisions is essential to the success of the program. 

This presentation will provide an overview of the more than 85 sites in the Legacy Management program, including 

a discussion of prior mission and associated environmental contamination issues; ramifications of the different regulatory 

and cleanup approaches taken during active remediation; and the challenges of the current long-term surveillance and 

maintenance mission. Additionally, the presentation will address lessons learned during the first eight years of the 

Legacy Management program in terms of risk management, public and stakeholder involvement, institutional controls, 

regulatory agreements and the issues associated with transition from active remediation to long-term surveillance and 

maintenance. Some of the case histories of how environmental management of these varied sites has evolved over time 

will be presented. 

2) 40270 – Lessons Learned in Planning the Canadian Nuclear Legacy Liabilities Program 

Michael Stephens, Sheila M. Brooks, Joan Miller, Robert Mason, AECL (Canada) 

In 2006, AECL and Natural Resources Canada (NRCan) began implementing a Nuclear Legacy Liabilities Program 

(NLLP) to manage the liabilities at AECL’s nuclear sites in Canada. <strong>The</strong>se liabilities include shutdown research and 

prototype power reactors, fuel-handling facilities, radiochemical laboratories, support buildings, radioactive waste 

storage facilities, and contaminated lands. <strong>The</strong> delivery of the Program is managed by a Liability Management Unit 

(LMU) within AECL. 

<strong>The</strong> NLLP comprises a large program of interlinked decommissioning, waste management and environmental 

restoration activities being executed at different sites, and by various technical groups as suppliers to the LMU. This 

paper describes the lessons learned in planning the “start-up” phase, which will conclude 2011 March, and the planning 

of the second phase, a 5-year program, which is currently being finalized. 

<strong>The</strong> initial “start-up” phase was planned by a small group of experts. Although effort was made to communicate the 

goals and overall strategy of the Program to the groups that would carry out the work, progress was slower than 

anticipated because AECL was ramping up from a minimal maintenance mode and the required increase in staff and 

technical resources was not sufficiently understood. Many of the projects being addressed within the Program are oneof-a-kind, 

and the essential base information on which to prepare detailed execution plans was not available to accurately 

plan the work. 

Internal reviews of the Program examined progress and identified several improvements to planning. <strong>The</strong>se 

included strengthening communications, conducting more advance planning of the interlinked activities, and building 

flexibility into the commitments made about activities that had yet to reach major decision points. 

<strong>The</strong> priorities for executing the required activities in the Program were set using criteria based on risks the liabilities 

presented to health, safety, the environment and to AECL’s business. In future, the decision criteria will also include the 

value gained for funds expended, and greater consideration will be given to mitigating risks to the execution of the 

Program. It was also determined that licensing strategies and processes need to be well-defined, and waste 

characterization methods and disposition pathways must be in place to deal with the wastes the Program will generate. 

Case studies will be presented in the paper to illustrate these lessons learned. 

<strong>The</strong> NLLP is funded by the Government of Canada through NRCan. 

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Abstracts 

3) 40292 – GIS-Based Decision Support System for the Former Semipalatinsk Test Site 

Maira Mukusheva, Sergey Baranov, National Nuclear Center (Kazakhstan) 

Termination of the nuclear tests and closure of the Semipalatinsk test site, taking into account the collapsed 

economy of the former Soviet Union brought to a situation when a huge area most part of which was contaminated with 

radioactive substances became open and uncontrolled providing a free access to the test site. Residents from the 

neighboring settlements used the opportunity. <strong>The</strong>y did not realize the hazard imposed during live-stock grazing, ferrous 

and non-ferrous metal scavenging, dismantling of buildings and structures located at the area. Grazing areas of some 

wintering grounds are located close to the nuclear tests areas. People living at the wintering ground and summer pastures, 

are occupied with grazing of cattle, sheep stock and drove of horses. Radioactive contamination of the site is caused by 

ground and above-ground nuclear explosions. In such a situation the main task was evaluation and prediction of 

radioecological situation, safety of economic activity. Application of geographic informational systems for analysis of 

radioecological process at radioactive contaminated nuclear test locations will make possible to process large amount of 

different information required for solution of problems related to the rehabilitation of contaminated areas. Efficiency, 

possibility to integrate ecological, social, economic, and management information on the unified geographic basis and 

combination with mathematic analysis tools, distinguish the geoinformational systems out of the existing applied 

programs. <strong>The</strong> universality determines the attractiveness of their application for a wide range of tasks. To evaluate and 

predict radioecological situation at the STS, a radioecological monitoring system (REMS) is developed on the basis of 

GIS-technologies. <strong>The</strong> system is designed to provide information, methodic and instrument support for preparation 

processes and management decision making in economic activity at the STS. To check the efficiency of the REMS and 

demonstrate applicability of GIS-technologies, risks calculation is performed for the STS population as part of analysis 

for consequences of ecosystems radioactive contamination. On the basis of analysis for references, field and laboratory 

experiments, main ways of pasture ecosystems contamination at STS are revealed and radionuclides are determined 

which make the most contribution to the additional radiation exposure to population carrying on economic activity in the 

region. Revised is the nature of agricultural production, schematic maps are constructed for location of economic agents. 

Using all available information, evaluation is carried out for actual condition and a prediction is made for possible 

negative consequences of radioactive contamination to the population of the former STS. 

4) 40218 – RFID Technology for Environmental Remediation and Radioactive Waste Management 

Hanchung Tsai, Yung Liu, ANL (USA); James Shuler, US DOE (USA) 

We have developed an advanced Radio Frequency IDentification (RFID) system capable of tracking and monitoring a wide 

range of materials and components for the nuclear industry – from fissionable stocks to radioactive wastes. <strong>The</strong> RFID technology has a 

number of advantages, such as enhanced safety and security, reduced personnel exposure to radiations, and improved inventory control. 

<strong>The</strong> sensors in the RFID tags monitor the state of health (e.g., temperature, shock, seal, dose, and dose rate) of the item and send out an 

alarm instantly when the sensor threshold is violated. Nonvolatile memories in the tag store data from sensors and event records, as 

well as manifest information if so desired. In irradiation tests, the tag electronics was confirmed to possess significant radiation 

resistance and, therefore, would yield a satisfactory service life. Long-life batteries and smart management circuitries permit the RFID 

tags to operate for up to 10 years without a battery replacement. <strong>The</strong> form factor of the tags can be modified to suit different container 

types. <strong>The</strong> read range can be up to 100 m, and no line-of-sight between the tagged items and the interrogator (reader) is necessary. 

With careful implementation, even a large-size processing or storage facility with a complex configuration can be monitored with a 

handful of readers in a network. In transportation, by using Global Positioning System (GPS) and satellite/cellular communication 

protocols, the locations and the conditions of the tagged containers can be continuously tracked. <strong>The</strong> RFID system also integrates 

Geographic Information Systems (GIS) technology, which uses information in preexisting geodatabases to generate and issue reports 

instantly to first responders for situation recovery in case of incidents. In stand-alone applications, the monitoring and tracking data are 

contained within the local control computer; with a secure Internet connection, multiple users can share the data in real time within the 

complex or beyond. As with the deployment of any new technology, overcoming the cultural resistance is part of the developmental 

process. With a strong institutional support and multiple convincing live demonstrations, the cultural resistance has been largely 

overcome. As a result, implementation of the RFID technology is taking place at several of U.S. Department of Energy installations for 

processing, storage, and transportation applications. 

5) 40181 – <strong>The</strong> Radioactivity of 3H in Metals by a High Temperature Furnace and a Liquid Scintillation Counter 

Hee Reyoung Kim, Geun Sik Choi, Sang Yun Park, Chang Woo Lee, Moon Hee Han, KAERI (Korea Rep.) 

<strong>The</strong> radioactivity of 3H of the metal samples from the nuclear sites was analyzed by using a commercialized high 

temperature furnace and a Liquid Scintillation Counter (LSC). <strong>The</strong> 3H activity of the sample was measured according to 

the duration of the high temperature combustion and the oxidation temperature. Basically, the recovery from the furnace 

was 90% for 3H and the LSC had a quenching efficiency of approximately 30 %. HNO3 was used as a trapping solution 

for 3H and the solution was cocktailed with a scintillator. <strong>The</strong> activity extracted from the sample was increased till the 

combustion time elapsed 60 minutes and the increasing rate was reduced continuously thereafter at 600 degree in 

centigrade whereas 80 % of radioactivity was extracted during the first 15 minutes at 900 degree in centigrade. Also, the 

pretreatment for the metal sample, which included a high temperature combustion and trapping, had the time required of 

at least four hours at 900 degree in centigrade. Finally, it was suggested that this high temperature combustion method 

could be applied to analyze the activity of the radioactive metal waste from the nuclear power plants. 

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Abstracts 

6) 40275 – Next Generation Waste Glass Melters in the U.S. DOE Waste Processing Program 

Steven P. Schneider, Gary Smith, US DOE (USA) 

<strong>The</strong> USA Department of Energy (U.S. DOE) Office of Environmental Management (EM) is evaluating alternative 

options for waste glass melting technologies. Specifically, DOEEM is assessing advanced melter technologies and 

developing a comprehensive research plan for next generation waste glass melter design and demonstration. Resolution 

of the USA nuclear waste legacy requires the design, construction and operation of large and technically complex 

one-of-a-kind processing facilities coupled to equally complex waste treatment and vitrification facilities. <strong>The</strong> loading of 

nuclear waste into glass and the glass production rates at U.S. vitrification facilities are limited by the current melter 

technology. Significant reductions in glass volumes for disposal and mission life are only possible with advancements in 

melter technology and glass formulations. Melters with higher throughput rate may shorten cleanup mission, in addition 

melters that allow for higher waste loading in glass may significantly reduce lifecycle costs. 

To help focus the next generation waste glass melter program, DOE-EM convened an international workshop to 

assess nuclear waste melter technologies and have used the melter workshop to help develop a comprehensive research 

plan for melter design and demonstration. <strong>The</strong> workshop included both oral presentations and discussion sessions from 

waste glass melter experts from around the world to assess the "state of the art" in melter technology and to lay the 

groundwork for a program plan that includes evolutionary changes to existing Joule-heated ceramic-lined liquid fed 

melters, as well as transformational melter technologies such as induction and hybrid-heated systems. At that workshop, 

representatives from many nations and international organizations (IAEA, China, France, Germany, India, Japan, Korea, 

Russia, UK, and the USA), universities (Catholic University of America, Missouri University of Science and 

Technology), and private companies (EnergySolutions, Kurion, URS) met to assess advanced melter technologies which 

helped the U.S. develop a comprehensive research plan for advanced waste glass melter design and demonstration with 

the goal of improved performance and reduced cost. <strong>The</strong> U.S. DOE Next Generation Waste Glass Melter Program is 

discussed in this paper. 


1) 40021 – Commercialization Project of Ulchin Vitrification� 

Hyun-jun Jo, Cheon-Woo Kim, KHNP (Korea Rep.); 

Tae-Won Hwang, Nuclear Engineering & Technology Institure (Korea Rep.) 

<strong>The</strong> Ulchin Vitrification Facility (UVF), to be used for the vitirification of low-and intermediate-level radioactive 

waste (LILW) generated by nuclear power plants (NPPs), is the world’s first commercial facility using Cold Crucible 

Induction Melter (CCIM) technology. <strong>The</strong> construction of the facility was begun in 2005 and was completed in 2007. 

From December 2007 to September 2009, all key performance tests, such as the system functional test, the cold test, the 

hot test, and the real waste test, were successfully carried out. <strong>The</strong> UVF commenced commercial operation in October 

2009 for the vitrification of radioactive waste. 

2) 40023 – Plasma Gasification/Vitrification of Wet ILW 

Gary Hanus, John Williams, Matt Zirbes, Phoenix Solutions Co. (USA) 

Magnox South Ltd has authorized a variety of decommissioning programmes to evaluate various technologies for 

timely and cost-effective remediation of a spectrum of waste streams resulting from the operation of their reactors. Of 

particular current interest is wet, intermediate level waste (ILW) in the form of solids, sludges and liquids. Hinkley Point 

A has over 137,000 litres of ILW organic cation resin containing significant quantities of radio-cesium (Cs). In 2009 

Phoenix Solutions Co was awarded a contract by Magnox South Ltd to demonstrate the effectiveness of thermal plasma 

treatment of this wet ILW resin waste stream. 

<strong>The</strong> objective of the wet ILW plasma treatment project was to feed organic resin sludge (containing a cesium 

surrogate) and a borosilicate glass frit simultaneously into a thermal plasma reactor to demonstrate the gasification of the 

organic content of the resin while capturing the cesium within a molten glass bath. A total of approximately 200 litres of 

glass product were produced. Approximately 5 metric tonnes of organic resin simulant were provided to Phoenix 

Solutions Co consisting of a mixture of 68% water / NaOH solution, 30% organic cation resin material with bound Cs 

and 2% other, including a small proportion of cesium in solution. <strong>The</strong> off-gas species from the plasma reactor exhaust 

were sampled by an independent, qualified sampling organization focusing on VOCs, HCl, NOx, SOx, CO, CO2, dioxins 

and furans as well as particulate levels. Glass samples were obtained and analyzed for crystalline inclusions, elemental 

identification, and viscosity characteristics at temperatures near the process melt conditions. <strong>The</strong> process temperature was 

maintained near 1000 ºC while the molten glass bath was held between 1100 and 1200 ºC. <strong>The</strong> processing produced 200 

litres of glass, 150 litres of which were successfully tapped from the plasma reactor into a standard Sellafield vitrified 

HLW disposal canister. 

This paper will describe the basic waste processing approach, the process hardware utilized, the process control 

features and the test results. Several trial tests were conducted, the longest of which processed over 2 metric tonnes of 

54

Abstracts 

dewatered resin waste together with 400 kg of glass. Further testing with improved process controls demonstrated an 

increase in cesium retention in the glass product. 

3) 40026 – Solidification Of Simulated Liquid Waste Of Primary Loop Resin Elution Process Of PWR 

Masamichi Obata, Michitaka Saso, Masaaki Kaneko, Nobuhito Ogaki, Taichi Horimoto, 

Toshiba corporation (Japan);Toshikazu Waki, <strong>The</strong> Kansai Electirc Power Co., Inc. (Japan) 

Primary loop resin waste is eluted by sulfuric acid in <strong>The</strong> Kansai Electic Company Mihama,Takahama and Oi 

nuclear power station. Waste solution from this elution process is planned to be solidified by cement. This study bring 

out a range of chemical composition and crud concentration of waste solution from this elution process, and examine the 

properties of alumina cement solidification process and solidified material. Test for sulfate ion, borate, lithium, 

ammonium ion was carried out. Volume reduction ratio of over 0.5 was archieved for 5 to 25wt% of sulfate ion and 

Abstracts 

SESSION L3: Nuclide Assay 

1) 40279 – Design of a facility for the automated receipt inspection and characterization of LILW using 

integrated non-destructive examination and assay techniques 

Stephen Halliwell, VJ Technologies Inc.� (USA) 

KHNP is constructing two repositories, located in Bonggil-ri, Yangbukmyun, Gyeongju-si, Gyeongbuk, Korea, for 

Low & Intermediate Level Radwaste. (LILW). <strong>The</strong> waste received is required to be inspected and characterized prior to 

being placed into storage at the repository. This paper describes the design of an integrated system of non destructive 

examination (xray), and non destructive assay (radiometric) instrument systems, waste package integrity checking 

systems, waste handling mechanisms and data management and archival provisions, which comprise the Waste Receipt 

and Inspection Facility (WRIF). <strong>The</strong> WRIF is designed to uncap transport containers, unload waste drums from 

containers, safely examine the drums, to ensure they meet the waste acceptance criteria established by the LILW facility. 

<strong>The</strong> drums are then loaded into waste disposal containers for transfer to the repository. <strong>The</strong> WRIF also receives larger 

waste containers, which are characterized in a separate section of the WRIF, before being transferred to the repository. 

Remote control and handling of the waste is a design requirement, as is the management, processing archival and 

retrieval system for all data generated during waste characterization. 

Importance of this work 

Remotely controlled characterization of LILW in drums and large boxes, by the integrated system of 

non-destructive and non-intrusive techniques in the WRIF, will minimize the risk of operator dose uptake, increase 

operator safety, reduce the generation of additional waste, and save time and cost within the waste management process. 

<strong>The</strong> data management, processing, archival and retrieval system will insure the data quality and integrity is maintained 

over the required period of time, and that particular data can be easily and reliably accessed at any time. 

2) 40167 – Feasibility Study on the Nuclide Analysis of the Radwaste Drum Using the Spectrum to Dose 

Conversion Factor 

Young-Yong Ji, Dae-Seok Hong, Tae-kuk Kim, Woo-Seog Ryu, KAERI (Korea Rep.) 

<strong>The</strong>re are several methods for non-destructive assay of a radwaste drum that are based on the gamma ray scanning 

and the in-situ objet counting system. Although these methods can be processed using high resolution, shielded detector 

and specific correction techniques to quantitatively analyze the nuclides in a drum, time and cost constraints compared 

with their accuracy dictate the use of simpler method to apply. Dose to curie (DTC) conversion method can simply and 

easily provide a reasonable estimate of the nuclide inventory in a radwaste drum. <strong>The</strong> measured dose rate as well as the 

relative abundance of gamma nuclides in a drum is a very important factor to be appreciated at the DTC conversion 

method because of the direct linearity between the measured dose rate and the gamma emitters in a drum. <strong>The</strong> dose rate is 

directly measured with the field detector. However, the relative abundance of gamma nuclides in a drum to be assayed is 

determined form an indirect measurement using the material balance by the waste stream. <strong>The</strong> uncertainty of the nuclide 

inventory of the assayed drum from the DTC conversion method could be increased because of the different detection 

mechanism between the dose rate and the relative abundance of gamma emitters in a drum. Unfortunately, that expands 

the limitation of using the DTC method. It is, therefore, necessary to find out a suitable measurement method of which 

two variables could be obtained from the drum to be assayed at once. This method could be realized by using the dose 

conversion factor, which has been widely using in the field of the environmental radiation measurement. <strong>The</strong> dose rate 

from a drum could be directly calculated from the measured gamma ray spectra by using the dose conversion factor, and 

also, the relative abundance of gamma nuclides could be easily obtained from the net count peaks in the spectra. In this 

study, the dose conversion factor for 3”?x3” NaI(Tl) scintillation detector around lead shield with the thickness of about 

3 cm was calculated by a MCNP code. <strong>The</strong> experimental verification for using this dose conversion factor was performed 

by using the simulated drum that has several holes for locating a standard source. 

3) 40255 – Portable Non-Destructive Assay Methods for Screening and Segregation of Radioactive Waste 

Alan Simpson, Martin Clapham, Stephanie Jones, Randy Lucero, Pajarito Scientific Corporation (UK) 

Significant cost-savings and operational efficiency may be realised by performing rapid nondestructive 

classification of radioactive waste at or near its point of retrieval or generation. <strong>The</strong>re is often a need to quickly 

categorize and segregate bulk containers (drums, crates etc.) into waste streams defined at various boundary levels (based 

on its radioactive hazard) in order to meet disposal regulations and consignor waste acceptance criteria. 

Recent improvements in gamma spectroscopy technologies have provided the capability to perform rapid in-situ 

analysis using portable and hand-held devices such as battery-operated medium and high resolution detectors including 

lanthanum halide and high purity germanium (HPGe). Instruments and technologies that were previously the domain of 

complex lab systems are now widely available as touch-screen “off-the-shelf” units. Despite such advances, the task of 

waste stream screening and segregation remains a complex exercise requiring a detailed understanding of programmatic 

requirements and, in particular, the capability to ensure data quality when operating in the field. This is particularly so 

when surveying historical waste drums and crates containing heterogeneous debris of unknown composition. <strong>The</strong> most 

56

Abstracts 

widely used portable assay method is based upon far-field High Resolution Gamma Spectroscopy (HRGS) assay using 

HPGe detectors together with a well engineered deployment cart (such as the PSC TechniCART technology). 

Hand-held Sodium Iodide (NaI) detectors are often also deployed and may also be used to supplement the HPGe 

measurements in locating hot spots. Portable neutron slab monitors may also be utilised in cases where gamma 

measurements alone are not suitable. 

Several case histories are discussed at various sites where this equipment has been used for in-situ characterization 

of debris waste, sludge, soil, high activity waste, depleted and enriched uranium, heat source and weapons grade 

plutonium, fission products, activation products, americium, curium and other more exotic nuclides. <strong>The</strong> process of 

acquiring and analyzing data together with integration of historical knowledge to resolve and delineate waste streams (for 

example between low-level waste and transuranic waste) is described. 

4) 40093 – Alpha radioactivity monitor using ionized air transportation for large size uranium waste (1) - Large 

measurement chamber and evaluation of detection performance – 

Susumu Naito, Shuji Yamamoto, Miikio Izumi, Yosuke Hirata, 

Yukio Yoshimura, Tatsuyuki Maekawa,Toshiba Corporation (Japan) 

Massive amounts of waste contaminated with alpha-radioactive uranium have accumulated in back-end facilities of 

the nuclear fuel cycle. In order to dispose of it adequately, it is necessary to classify such waste according to its so-called 

clearance level. However, when measuring alpha radioactivity using a conventional survey meter, one must hold it as 

close as possible to the surface, due to the relatively short flight range of alpha particles in air. As a result, measuring a 

large amount of diverse waste takes a long time and involves high labor costs. <strong>The</strong>refore, a new efficient alpha 

measurement technique is strongly desired. To satisfy this demand, we have been developing an alpha radioactivity 

monitor based on the principle of alpha radioactivity measurement using ionized air transportation. <strong>The</strong> measurement 

principle is as follows. Air is ionized by alpha particles emitted from the waste. <strong>The</strong> produced ions are transported to an 

ion sensor with the air flow produced by a blower, where the number of ions is measured as an electric current (ion 

current). <strong>The</strong> alpha radioactivity is then evaluated from the ion current value. This indirect measurement method is very 

efficient, because the ions produced near the entire surface of the waste can be detected both all at once and at a distant 

position from the waste. In previous work, we developed a prototype monitor with an about 1000 mm cubic measurement 

chamber to measure the cut waste. However, in a survey of target waste, we found that it is desired to measure not only 

the cut waste but also the lengthy waste such as uncut cylinders. <strong>The</strong>refore, we developed an alpha radioactivity monitor 

with a long and large measurement chamber (effective sizes: 500 mm x 900 mm x 3200 mm) for the long and large 

cylindrically-shaped waste (maximum size: 300 mm in diameter and 3000 mm in length, weight: 10 to 200 kg). We 

aimed

Abstracts 

6) 40111 – Preparation of Reference Materials on Radiochemical Analysis for Low-Level Radioactive Waste 

Generated from Japan Atomic Energy Agency 

Ken-ichiro Ishimori, Mikio Nakashima, Kuniaki Takahashi, Yutaka Kameo, JAEA (Japan) 

We have advanced the development of simple and rapid determination method for important radionuclides on safety 

assessment for disposal of low-level radioactive wastes generated from Japan Atomic Energy Agency (JAEA). In the 

radiochemical analyses of the radioactive wastes, it is necessary to manage accuracy and precision of determined values 

of radioactivity concentration using reference materials. However, since appropriate reference materials of radioactive 

waste are hardly available at the present state, the developments of laboratory-scale preparation methods are required to 

supply reference materials. In this work, we investigated preparation methods for the reference materials containing 

important nuclides and confirmed the validity of the prepared materials. Additionally a reference material for cemented 

liquid waste was also prepared. --Solidified product containing alpha-ray and gamma-ray emitting nuclides--In waste 

management in JAEA, non-metallic low-level radioactive solid wastes will be treated by plasma melting at the Advanced 

Volume Reduction Facilities (AVRF). In order to clarify optimum melting conditions of solidified products using a 

laboratory-scale electric furnace instead of plasma heating device, we conducted melting tests of a miscellaneous 

simulated solid waste in the presence of stable isotope tracers. Over 90% of Cs remained in the solidified product by 

keeping the basicity (CaO[wt%] / SiO2[wt%]) to be 0.05. Under the optimum melting conditions, we prepared reference 

materials containing alpha-ray (237Np, 241Am, and 244Cm) and gamma-ray (60Co, 137Cs, and 152Eu) emitting 

nuclides. <strong>The</strong> characteristics observed in SEM-EDX measurement and chemical durability against acids suggested that 

glass structure of the reference materials was almost same as that of solidified products produced by plasma melting. 

--Solidified product containing 14C or 36Cl-- Since 14C and 36Cl easily vaporize at a high temperature, it is difficult to 

remain the nuclides in solidified product on melting treatment with an electric furnace. First, melting conditions for 

solidified glasses containing N or Cl were optimized. We attempted a preparation method which produces 14C or 36Cl in 

the solidified glass using nuclear reaction 14N(n, p)14C or 35Cl(n, gamma)36Cl by thermal neutron irradiation. 

Reference materials containing 14C or 36Cl were successfully prepared by the proposed method. <strong>The</strong> radioactivity 

concentrations of the reference materials were evaluated from the developed simple and rapid determination method. 

From the results, it was confirmed that reference materials, which was useful for routine radiochemical analysis, could be 

successfully prepared on the present preparation methods. 

SESSION H4: National and <strong>International</strong> Programs (2) 

1) 40213 – U.S. NRC Integrated Spent Fuel Management Plan 

Catherine Haney, Shawn Smith, US NRC (USA) 

<strong>The</strong> U.S. Nuclear Regulatory Commission (NRC) is developing an integrated plan for regulating the interrelated 

activities that are involved in the management of spent fuel and high-level waste. Treating the system as a whole is 

essential for several reasons: (1) decisions made about one component or activity of the waste management system could 

significantly affect other components (e.g., a decision made in isolation could inadvertently impact alternatives to the 

system as a whole); (2) treating waste management as a system is a more efficient and effective way to determine 

priorities, logically complete activities or to appropriately deal with unexpected situations that may arise with 

first-of-a-kind programs like reprocessing or alternative waste disposal options; and (3) viewing the system as a whole 

avoids gaps and unnecessary duplication in regulations leading to more effective and efficient development and 

application of regulatory oversight. 

Near-term flexibility is a key consideration because the national policy in the USA appears to be changing and will 

likely remain in flux for some time. For example, the Secretary of Energy has convened a commission to conduct a 

comprehensive review of the policies for managing the back end of the nuclear fuel cycle. NRC activities will be 

informed by this and other relevant developments as it develops its plans. However, integration is essential regardless of 

the direction of national policy, as the NRC needs to remain flexible and agile under a range of policy outcomes. This 

presentation will describe key aspects of the plan and how the plan will be implemented. 

2) 40116 – Regulatory Research for Geological Disposal of High-level Radioactive Waste in Japan 

Shinichi Nakayama, JAEA (Japan); Yoshio Watanabe, AIST (Japan); Masami Kato, JNES (Japan) 

<strong>The</strong> Nuclear and Industrial Safety Agency of the Ministry of Economy, Trade and Industry (NISA) has renewed its 

regulatory role and supporting research needs, taking into consideration recent circum-stances of radioactive waste 

management in Japan. In October 2009, a technical support organization of the Japan Nuclear Energy Safety 

Organization (JNES) released its independent five-year research plan, “Regulatory Support Research Plan on Radioactive 

Waste Management 2010-2014”, in cooperation with the research institutes of Japan Atomic Energy Agency (JAEA) and 

the National Institute of Advanced In-dustrial Science and Technology (AIST). <strong>The</strong> plan covers low- through high-level 

radioactive waste management. <strong>The</strong> geological disposal research plan and the future research activities are outlined in 

this paper. Japan’s nuclear regulation law, “Law for Regulation of Nuclear Source Materials, Nuclear Fuel Ma-terials and 

58

Abstracts 

Nuclear Reactors”, was amended in 2007 to address the safety of geologic disposal of high-level radioactive waste. NISA 

announced its involvement or supervision not only for the licensing application but during the site selection process. Two 

major research areas required by NISA have been identified: studies to review the validity of preliminary survey results 

during the site selection, and studies to review the safety assessment for a licensing application. <strong>The</strong>se research areas are 

aimed at 1) developing “safety indicators” to judge the adequacy of site investigation results presented by the operator, 2) 

compiling basic requirements of safety design and safety assessment needed to make a technical evaluation of the license 

application, as well as developing safety indicators for objective evaluation, and 3) developing an inde-pendent safety 

assessment methodology. In addition, NISA plans to periodically issue the report, “<strong>The</strong> Regulatory Research Report on 

Geological Disposal” consistent with the operator’s technical report sched-ule. This report would be intended to confirm 

the regulatory status of the program as well as strengthen-ing the competence of NISA as a regulatory body. JNES 

launched safety studies on geological disposal in its establishment year in 2003. JAEA and AIST joined as regulatory 

support research institutes in 2005. In October 2007, the three parties signed an agreement of cooperative study on 

geological disposal, which enhanced joint studies, as well as exchanges of staff, data, and results. One of the ongoing 

joint studies has been aimed at investigating regional-scale hydrogeological modeling using JAEA’s Horonobe 

Underground Research Center. <strong>The</strong> three parties have begun to discuss expanding the joint studies and the agreement 

areas in response to the new five-year re-search plan. 

3) 40280 – Recent Developments and Trends on Requirements Management Systems 

Satoru Suzuki, Hiroyoshi Ueda, Kiyoshi Fujisaki, Katsuhiko Ishiguro, Hiroyuki Tsuchi, NUMO (Japan); 

Stratis Vomvoris,Irina Gaus, Nagra (Switzerland) 

Although the management of requirements for the development of geologic repository systems has been practiced in 

all radioactive waste disposal programs from the beginning, the systematic management through IT-based requirement 

management systems (RMS) has a younger history. Traditionally requirements have been gathered in various documents 

and quite often the rationale for design choices is difficult to trace. Thanks to the development of RMS in other industries 

(software development, aerospace), many radioactive waste disposal programs have recently been able to set-up and 

tailor RMS to their needs quickly. In a recent international meeting, five radioactive waste disposal organizations 

(NUMO/Japan; NAGRA/Switzerland; ONDRAF/NIRAS/Belgium; POSIVA/Finland and SKB/Sweden) have discussed 

the status and developments of RMS in their respective programs. <strong>The</strong> majority have already implemented an IT-based 

system, or, are testing and developing such systems. <strong>The</strong> level of detail of requirements depends on the stage of the 

program. Those approaching the license application have integrated all components of the repository concept, including 

the processes for the operational phase. It was recognized that the earlier you implement a systematic tracking of 

requirements and the decisions taken to satisfy these requirements, the easier it would be to implement an RMS over the 

whole duration of the geologic disposal program – in the order of 100 years. <strong>The</strong> documented transition from the general 

requirements (highest level) to the lower level ones – especially those related to the repository system concept or its 

components – can be then used to enable the assessment of potential changes at future stages of the system. Even if the 

host rock or, the repository concept has not been selected, a ‘dry’ run with assumed conditions can be very elucidating of 

the most useful set-up of the RMS. Requirements management is closely associated with the quality management system. 

Combining requirement and decision-tracking has been expressed as an explicitly goal for some programs. Caution was 

expressed regarding the expectations for the RMS being developed. <strong>The</strong>re is a risk that such systems are perceived as 

expert systems that can derive decisions, which then will be unquestionably accepted. It is nevertheless recognized that 

they can be of great help in communicating with the various stakeholders and with relative ease demonstrating how their 

requirements have been considered and satisfied with the proposed repository systems. Further efforts need to be 

undertaken to integrate the requirement management systems, and the processes that they represent, in the day-to-day 

operations of the organizations. Different organizational schemes are being considered, for example, the composition of 

the team that defines the requirements at the various levels and the function of its members within the technical and 

scientific program of the waste disposal organizations. First positive experiences of the latter were reported. 

4) 40228 – Development of Requirements Management System of NUMO and practical experience with 

development of the database contents 

Satoru Suzuki, Hiroyoshi Ueda, Katsuhiko Ishiguro, Hiroyuki Tsuchi, Kiyoshi Fujisaki, NUMO (Japan); 

Kiyoshi Oyamada, JGC Corporation (Japan); Shoko Yashio, Obayashi Corporation (Japan) 

Decision-making and work activities in the geological disposal program need to be implemented in such a way as to 

fulfill various requirements such as safety, practicality, quality and socio-economic aspects. Since a stepwise approach is 

applied for implementing the program, the number, weighting and specific nature of the requirements will change 

depending on the premises and constraints in each stage of implementation. Requirements management with a long-term 

perspective is therefore required for consistent implementation of the program. NUMO has developed a requirements 

management (RM) methodology that is suitable for the long-term, stepwise disposal program in Japan, as well as a 

supporting requirements management system (RMS) tool. <strong>The</strong> basic concept of the RMS was already presented at the 

last ICEM 2009. In this presentation, we will focus on practical experience with development of the database content for 

the RMS. <strong>The</strong> RM methodology was first applied in the HLW repository design work. Requirements for repository 

design were considered primarily from the viewpoint of post-closure safety and engineering feasibility. <strong>The</strong> repository 

design requirements are structured hierarchically to those assigned to the ‘post-closure safety concept’, the ‘required 

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Abstracts 

system functions’ and the ‘design requirements’. <strong>The</strong> post-closure safety concept, which is based on isolation by a 

multibarrier system, is further subdivided into fundamental safety functions and operational functions under the heading 

‘required system functions’. <strong>The</strong> fundamental safety functions are ‘post-closure containment’, ‘retardation/reduction of 

radionuclide migration’ and ‘length of migration pathway’ and they need to be satisfied in order to fulfill the safety 

concept. <strong>The</strong> aim of the operational functions is to realize these safety functions by providing a feasible and reliable 

operating system (e.g. welding facility, transport and emplacement). Finally, the lowest level – the design requirements - 

are defined for each system component, so that each component can fulfill the upper ranking ‘required system functions’. 

<strong>The</strong> system components are the engineered barrier system (metal overpack and the bentonite buffer) and the natural 

barrier (the host formation and its surroundings). In the NUMO RMS tool, four elements -‘Decision/Work (D/W)’, 

‘Requirements (R)’, ‘Conditions (C)’ and ‘Arguments (A)’ are used for describing the RM information. <strong>The</strong> design 

requirements discussed above are recorded in the ‘R’ element and attributed to the design work of each component of the 

‘D/W’ elements. <strong>The</strong> ‘C’ elements record information such as the site environment, which in turn affects the D/W and R 

elements. ‘A’ elements record the synthesis of evidence that fulfills the requirements. <strong>The</strong> database content for these four 

elements will be defined in advance before initiating the design work in each stage of the disposal program. 

5) 40231 – Application of Lifecycle Management to Design of the UK Geological Disposal Facility 

Henry O'Grady, Malcolm Currie, Parsons Brinckerhoff (UK); 

Philip Rendell, Radioactive Waste Management Directorate (UK) 

<strong>The</strong> Radioactive Waste Management Directorate (RWMD) of the UK’s (UK) Nuclear Decommissioning Authority 

(NDA) has been charged with the responsibility for design and delivery of a Geological Disposal Facility (GDF) for High 

Level Waste / Intermediate Level Waste in accordance with government policy. Over the next few years, the design of 

the GDF will be developed and the organisation built up to provide the necessary design and delivery capability. In the 

short term, this capability must also be demonstrated to UK government regulators in order to gain approval for nuclear 

operations. As part of this process, RWMD have developed a staged approach to engineering design, which addresses the 

overall lifecycle of the GDF in terms of seven phases, from initial concept development through to operation and, finally, 

closure. Each phase finishes with a formally defined milestone (a “gate”) comprising a technical review and a specific set 

of engineering deliverables. <strong>The</strong> phases and milestones have been built up from a number standard approaches described 

in open literature, which have then been crafted to address the specific needs of the GDF. Roles and responsibilities are 

also considered, along with the interface issues between various functional groups both within and outside of the 

engineering sphere. <strong>The</strong> process also incorporates a number of good practices based on the authors’ experiences, such as 

requirements management, progressive assurance and the use of architectural frameworks. As the lifecycle of the GDF 

will extend over decades, the process uses a modular approach which will permit it to evolve to meet the changing needs 

of the project, the organisation and the regulatory process. <strong>The</strong> process is intended to help the engineering effort provide 

timely support to, the higher-level needs of the overall project and the community engagement programmes, as well as 

with the activities of non-engineering functions. <strong>The</strong> process also provides for the management of risk by ensuring that 

the requirements, designs, risk registers and detailed procedures are accepted before further funding is released. This 

paper describes the background to the UK GDF programme, the organisational issues associated with the RWMD’s 

evolving role, the relationship between the top-level UK Government Managing Radioactive Waste Safely programme 

and the RWMD engineering lifecycle, the formal reviews, the milestones and the overall contribution this makes to 

RWMD organisational development and UK regulatory approval. 

SESSION D2: Dismantling and Decontamination 

1) 40036 – AREVA NP: Experience in Dismantling and Packaging of Pressure Vessel and Core Internals 

Peter Pillokat, Jan Hendrik Bruhn, AREVA NP GmbH (Germany) 

AREVA NP GmbH, German Regional Sector of French Nuclear Company AREVA is proud to look back on 

versatile experience in successfully dismantling nuclear components. After performing several minor dismantling 

projects and studies for nuclear power plants, AREVA NP completed the order to dismantle all remaining reactor 

pressure vessel internals at German boiling water reactor Wuergassen NPP in October ´08. During the onsite activities 

about 121 tons of steel where successfully cut and packed under water into 200l- drums, as the dismantling was 

performed partly in situ and partly in an underwater working tank. AREVA NP deployed a variety of different cutting 

techniques such as band sawing, milling, nibbling, compass sawing and water jet cutting throughout this project. After 

successfully finishing this task, AREVA NP is currently dismantling the cylindrical part of the Wuergassen Pressure 

Vessel. During this project approxi-mately 320 tons of steel are cut and packaged for final disposal, as dismantling is 

mainly performed by on air use of water jet cutting with vacuum suction of abrasive and kerfs material. <strong>The</strong> main clue 

during this assignment is the logistic challenge to handle and convey cut pieces from the pressure vessel to the packing 

area. For this an elevator is installed to transport cut segments into the turbine hall, where a special housing is built for 

final storage conditioning. At the beginning of ‘07 another complex dismantling project of great importance was acquired 

by AREVA NP. <strong>The</strong> contract included dismantling and conditioning for final storage of all the RPV Internals at German 

pressurized water reactor Stade NPP. Very similar cutting techniques turned out to be the proper policy to cope this task. 

60

Abstracts 

Onsite activities took place in up to 5 separate working areas including areas for post segmentation and packaging to 

perform optimized parallel activities. All together about 85 tons of core internals where successfully dismantled at Stade 

NPP until September ´09. To accomplish the best possible on-site performance and to achieve a minimization of the 

applied collective dose rate, each onsite activity was previously planned in detail and personnel exercised each task at 

original size mockups under most realistic onsite conditions. Planning was especially focused on an optimized size 

minimization and packaging concept to reduce the number of filled waste packages. <strong>The</strong> segmentation of components 

strictly followed a sophisticated cutting and packaging concept developed under consideration of possible cutting 

techniques, the resulting geometry and logistical conditions. <strong>The</strong>refore segments were post processed by hydraulical 

press and band saw in order to minimize their volume, were applicable. 

2) 40102 – Study on evaluation models of management data for decommissioning of Fugen 

Yuji Shibahara, Masanori Izumi, Takashi Nanko, Mitsuo Tachibana, Tsutomu Ishigami, JAEA (Japan) 

In the Fugen nuclear power plant (FUGEN), the dismantling of equipments in the turbine building has started in 

2008, and the dismantling of equipments around the reactor is scheduled around in 2015. To evaluate the management 

data on this dismantling of equipments around reactor appropriately, it is very important to study whether the 

conventional evaluation models have the applicability for FUGEN or not. Thus, the management data on the dismantling 

of equipments in 3rd/4th feedwater heater room conducted in 2008 was calculated with the conventional evaluation 

models. <strong>The</strong> conventional evaluation models were made by data obtained from Japan Power Demonstration Reactor 

(JPDR) decommissioning program. It was found that there were large differences between the calculated values and the 

actual data. For finding the cause in the difference between them, the dismantling of equipments in 3rd/4th feedwater 

heater room was divided into three processes: i) the preparation process, ii) the dismantling process, and iii) the clean-up 

process. In the both process of preparation and clean-up, the calculated values were smaller than the actual data. <strong>The</strong>se 

were mostly caused by the plant scale difference between JPDR and FUGEN, because the conventional evaluation 

models were built by analyzing the actual data on the decommissioning of JPDR which is smaller than FUGEN. <strong>The</strong> 

primary expression depending on the area of working space was built as new evaluation models for the preparation and 

clean-up processes. In the dismantling process, on the other hand, it was found that there were characteristic differences 

in the dismantling of feedwater heater as follows: 1) the calculated values were significantly larger than the actual data, 

2) the actual data for the dismantling of 3rd feedwater heater was larger than that of 4th one, though these equipments 

were almost same weight. It was found that these were brought by the difference in the descriptions of dismantling of 

feedwater heaters, and the new evaluation models reflecting the descriptions of dismantling were built for the appropriate 

evaluation of the management data. <strong>The</strong> calculated values with the new evaluation models for each process showed the 

good agreement with the actual data. In this report, study on evaluation models of management data for dismantling of 

equipments in the feedwater heater room will be described. 

3) 40083 – CORD Decontamination Technologies for Decommissioning - A Comprehensive Approach Based on 

Over 30 Years Experience 

Christoph Stiepani, AREVA NP GmbH (Germany) 

Decontamination prior to Decommissioning and Dismantlement is imperative. Not only does it provide for 

minimization of personnel dose exposure but also maximization of the material volume available for free release. Since 

easier dismantling techniques in lower dose areas can be applied, the licensing process is facilitated and the scheduling 

and budgeting effort is more reliable. <strong>The</strong> most internationally accepted approach for Decontamination prior to 

Decommissioning projects is the Full System Decontamination (FSD). FSD is defined as the chemical decontamination 

of the primary cooling circuit, in conjunction with the main auxiliary systems. 

AREVA NP has long-term experience with Full System Decontamination for return to service of operating nuclear 

power plants as well as for decommissioning after shutdown. Since 1976, AREVA NP has performed over 500 

decontamination applications and, from 1986, AREVA NP has performed Decontaminations prior to Decommissioning 

projects which comprise virtually all NPP designs and plant conditions: 

- NPP designs: HPWR, PWR, and BWR by AREVA, Westinghouse, ABB and GE 

- Decontaminations performed shortly after final shutdown or several years later, and even after re-opening Safe 

Enclosure 

- Gamma / Alpha inventory 

- Main Coolant chemistry (e.g., with and without Zn injection during operation) 

Fifteen Decontamination prior to Decommissioning Projects have been performed successfully to date. <strong>The</strong> 

sixteenth FSD for Decommissioning at the French NPP Chooz A is now in the detailed engineering phase with onsite 

application scheduled for Fall 2010. 

This paper will describe the AREVA NP Decontamination Concept for Decommissioning (DCD) and present 

highlights of previous FSDs performed prior to decommissioning using the CORD / AMDA technology. 

4) 40007 – Chemical Decontamination for Decommissioning (DFD) and DFDX 

Ronald Morris, Westinghouse Electric Company (USA) 

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Abstracts 

DFD is an acronym for the “Decontamination for Decommissioning” process developed in 1996 by the Electric 

Power Research Institute (EPRI). <strong>The</strong> process was designed to remove radioactivity from the surfaces of metallic 

components to allow these components to be recycled or free-released for disposal as non-radioactive. DFD is a cyclic 

process consisting of fluroboric acid, potassium permanganate and oxalic acid. <strong>The</strong> process continues to uniformly 

remove base metal once oxide dissolution is complete. 

<strong>The</strong> DFD process has been applied on numerous components, sub-systems and systems including the reactor 

systems at Big Rock Point and Maine Yankee in the USA, and the Jose Cabrera (Zorita) NPP in Spain. <strong>The</strong> Big Rock 

Point site has been returned to Greenfield and at Maine Yankee the land under the license was reduced for an 

Independent Spent Fuel Storage Installation (ISFSI). In the upcoming months, the Zorita NPP in Spain will initiate 

dismantlement and decommissioning activities to return the site to a non-nuclear facility. 

<strong>The</strong> development of the EPRI DFD process has been an on-going evolution and much has been learned from its use 

in the past. It is effective in attaining very high decontamination factors; however, DFD also produces secondary waste in 

the form of ion exchange resins. This secondary waste generation adds to the decommissioning quota but can be 

improved upon at a time when radioactive waste storage at nuclear facilities and waste disposal sites is limited. 

To reduce the amount of secondary waste, EPRI has developed the DFDX process. This new process is an 

enhancement to the DFD process and produces a smaller amount of metallic waste rather than resin waste; this reduction 

in volume being a factor of ten or greater. Electrochemical ion exchange cells are the heart of the DFDX system and 

contain electrodes and cation ion exchange resin. It has been used very successfully in small system applications and the 

next evolution is to design, build and implement a system for the chemical decontamination for decommissioning of 

larger reactor systems and full system decontamination (FSD). 

<strong>The</strong> purpose of this paper will be to provide a reference for those planning future chemical decontaminations for 

plant decommissioning. It will be based on actual experience from the work that has been performed to date and the 

planned development of the DFDX process. 

5) 40127 – Methods for Calculation and Optimisation of Personnel Exposure during Planning of 

Decommissioning of Nuclear Installation 

Marek Vasko, Ivan Rehak, DECOM, a.s. (Slovakia); Vladimir Daniska, DECONTA, a.s. (Slovakia); 

Vladimir Necas, Slovak University of Technology in Bratislava (Slovakia) 

Assessment of personnel exposure is one of safety issues being evaluated within decommissioning plans. It is 

required to show an impact of planned decommissioning activities to personnel and demonstrate even at the stage of 

decommissioning planning that the personnel exposure will be minimized in line with ALARA principle. <strong>The</strong> paper 

presents a methodology for evaluation of personnel exposure developed within standardized decommissioning costing 

code OMEGA. It deals with a methodology for calculation of external and internalpersonnel exposure based on 

calculated individual manpower components for each profession within working group and relevant dose rates, the dose 

rate from equipment, average dose rate in the rooms and background dose rate, as well as air volume radioactivities of 

radionuclides as they are recorded in facility inventory database. It also deals with the methodology for evaluation and 

optimization of the individual effective dose for individual members of working group based on their profession. 

Developed methodology enables optimization of deployment for individuals assigned to given professions within 

deployed working group in order to keep the individual effective dose within stipulated limits. 

SESSION D3: Planning 

1) 40129 – Program Change Management During Nuclear Power Plant Decommissioning 

Mike Naughton, Sean Bushart, Karen Kim, EPRI (USA) 

Decommissioning a nuclear power plant is a complex project. <strong>The</strong> project involves the coordination of several 

different departments and the management of changing plant conditions, programs, and regulations. As certain project 

Milestones are met, the evolution of such plant programs and regulations can help optimize project execution and cost. 

This paper will provide information about these Milestones and the plant departments and programs that change 

throughout a decommissioning project. <strong>The</strong> initial challenge in the decommissioning of a nuclear plant is the 

development of a definitive plan for such a complex project. EPRI has published several reports related to 

decommissioning planning. <strong>The</strong>se earlier reports provided general guidance in formulating a Decommissioning Plan. 

This Change Management paper will draw from the experience gained in the last decade in decommissioning of nuclear 

plants. <strong>The</strong> paper discusses decommissioning in terms of a sequence of major Milestones. <strong>The</strong> plant programs, associated 

plans and actions, and staffing are discussed based upon experiences from the following power reactor facilities: Maine 

Yankee Atomic Power Plant, Yankee Nuclear Power Station, and the Haddam Neck Plant. Significant lessons learned 

from other sites are also discussed as appropriate. Planning is a crucial ingredient of successful decommissioning projects. 

<strong>The</strong> development of a definitive Decommissioning Plan can result in considerable project savings. <strong>The</strong> decommissioning 

plants in the U.S. have planned and executed their projects using different strategies based on their unique plant 

circumstances. However, experience has shown that similar project milestones and actions applied through all of these 

projects. This allows each plant to learn from the experiences of the preceding projects. As the plant transitions from an 

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operating plant through decommissioning, the reduction and termination of defunct programs and regulations can help 

optimize all facets of decommissioning. This information, learned through trial in previous plants, can be incorporated 

into the decommissioning plan of future projects so that the benefits of optimization can be realized from the beginning 

of the projects. This process of the collection of information and lessons learned from plant experiences is an important 

function of the EPRI Decommissioning Program. 

2) 40245 – Status of the Support Researches for the Regulation of Nuclear Facilities Decommissioning in Japan 

Yusuke Masuda, Yukihiro Iguchi, Satoru Kawasaki, Masami Kato, JNES (Japan) 

In Japan, 4 nuclear power stations are under decommissioning and some nuclear fuel cycle facilities are expected to 

be decommissioned in the future. On the other hand, the safety regulation of decommissioning of nuclear facilities was 

changed by amending law in 2005. An approval system after review process of decommissioning plan was adopted and 

applied to the power stations above. In this situation, based on the experiences of the new regulatory system, the system 

should be well established and moreover, it should be improved and enhanced in the future. Nuclear Industry and Safety 

Agency (NISA) is in charge of regulation of commercial nuclear facilities in Japan and decommissioning of them is 

included. Japan Nuclear Energy Safety Organization (JNES) is in charge of technical supports for NISA as a TSO 

(Technical Support Organization) also in this field. As for decommissioning, based on the needs in terms of regulation, 

JNES has been continuing research activities from October 2003, when JNES has been established. Considering the 

“Prioritized Nuclear Safety Research Plan (August 2009)” of the Nuclear Safety Commission of Japan and the situation 

of operators facilities, “Regulatory Support Research Plan between FY 2010-2014” was established in November 2009, 

which shows the present regulatory needs and a research program. This program consists of researches for 1. review 

process of decommissioning plan of power reactors, 2. review process of decommissioning plan of nuclear fuel cycle 

facilities, 3. termination of license at the end of decommissioning and 4. management of decommissioning waste. For the 

item 1, JNES studied safety assessment methods of dismantling, e.g. obtaining data and analysis of behavior of dust 

diffusion and risk assessment during decommissioning, which are useful findings for the review process. For the item 2, 

safety requirements for the decommissioning of nuclear fuel cycle facilities was compiled, which will be used in the 

future review. For the item 3, measuring method, release procedure and analysis code for the site release were studied for 

the establishment of the license termination process in the future. From FY 2010, based on the new plan, we have started 

the researches for the standardization of review process of decommissioning plan for power reactors and nuclear fuel 

cycle facilities, establishing the process and criteria of license termination and appropriate method of management of 

decommissioning waste based on the waste form confirmation process. 

3) 40136 – Decommissioning Costing Approach Based on the Standardised List of Costing Items; Lessons 

Learnt 

Vladimir Daniska, Frantisek Ondra, Peter Bezak, DECONTA, a.s. (Slovakia); 

Ivan Rehak, Marek Vasko, Jozef Pritrsky, DECOM a.s. (Slovakia) ; 

Matej Zachar, Vladimir Necas, Slovak University of Technology in Bratislava (Slovakia) 

<strong>The</strong> document “A Proposed Standardised List of Items for Costing Purposes” was issues in 1999 by OECD/NEA, 

IAEA and European Commission for promoting the harmonisation in decommissioning costing. It is a systematic list of 

typical decommissioning activities classified into chapters 01 to 11 with three hierarchical numbered levels. Cost groups 

are defined in chapter 12 for presenting cost for each activity. In this way the document is the standardised matrix of 

decommissioning activities and cost groups with unambiguous meaning of items. Knowing what is behind the items of 

the standardised cost structure makes the comparison of cost for various decommissioning projects transparent. <strong>The</strong>re are 

two principal approaches for use of the standardised cost structure. First approach converts the cost items from specific 

cost structures into the standardised cost structure for the purpose of cost presentation. Second approach uses the 

extended standardised cost structure as the calculation structure; the cost data have the standardised format directly. 

Several advantages can be identified in this approach. <strong>The</strong> paper presents general aspects of the costing methodology 

based on the standardised cost structure and lessons learnt from last ten years of implementation of the standardised cost 

structure as the cost calculation structure in the computer code OMEGA. <strong>The</strong> paper presents following aspects: 

• general principles of decommissioning costing in current costing methodologies based on information on costing 

methodologies available, 

• standardised cost structure, methods of its use in costing, cost calculation structures based on the standardised 

cost structure, 

• principles of cost calculation methodology based on the standardised cost structure, input data for the 

methodology, cost calculation process, management of calculated data 

• interactions of main cost structures involved in standardised costing: standardised cost calculation structure, 

work breakdown structure of a decommissioning project, relations to cost accounting structures 

• main features of the computer code OMEGA: implementation of the standardised cost structure as the cost 

calculation structure, management of the material and radioactivity flow in the decommissioning process, 

generation of Gantt charts for managing of the decommissioning projects, data feedback from real processes, 

links to cost accounting systems 

• international activities for promoting of the standardised cost structure and for its upgradin 

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4) 40290 – <strong>The</strong> Outline of Decommissioning Plan for Hamaoka Nuclear Power Station Unit-1 and Unit-2 

Yoshifusa Fukuoka, Chubu Electric Power Co., Inc. (Japan) 

Hamaoka Nuclear Power Station's Unit-1 and Unit-2 ended their operation on January 30, 2009. <strong>The</strong> Unit-1 and 2 

will be dismantled and removed, and doing this requires establishing a nuclear reactor facility decommissioning plan, and 

getting the approval of the national government. <strong>The</strong>ir decommissioning procedure commenced after the submission of 

the "Application for the approval of the decommissioning plan for Hamaoka Nuclear Power Station Unit-1 and 2" to the 

Minister of Economy, Trade and Industry on June 1, 2009 , and subsequent approval for the application on November 18, 

2009. ?<strong>The</strong> application includes an overall plan for dismantling reactor facilities safely and surely, a description of tasks 

to be performed during the period Chubu Electric is preparing to dismantle the facilities in the coming years and safety 

assurance measures, among other information. ?According to the plan, the decommissioning of Unit-1 and Unit-2 is to be 

completed by the end of FY2036. <strong>The</strong> 28-year schedule is divided into four phases. ?Phase 1?Preparation Stage, 

FY2009 ? FY2014 Transporting and transferring nuclear fuel out of the plants, investigating the status of contamination, 

conducting system decontamination, and dismantling / removing facilities and equipment outside RCA ?Phase 

2?Dismantling and Removal Stage for Reactor Zone Peripheral Facilities, FY2015 ? FY2022 Dismantling / removing 

reactor zone peripheral facilities, safely storing dismantlement debris, and installing facilities for processing 

dismantlement debris ? Phase 3?Dismantling and Removal Stage for Reactor Zones, FY2023 ? FY2029 Dismantling and 

removing the reactor zone (covering the reactor vessel, core support structures, and radiation shields surrounding the 

reactor vessel) ?Phase 4?Dismantling and Removal Stage for Building Structures, FY2030 ? FY2036 Removing 

radioactive materials inside plant buildings, and dismantling / removing the buildings. 

Of waste to be generated in decommissioning work at Unit-1 and Unit-2, low-level radioactive waste accounts for 

approx. 17,000 tons . Low-level radioactive waste is sorted according to the types of radioactive substances contained or 

the level of radiation based on laws and regulations, and appropriately put to underground disposal based on the 

classifications. Specific methods for disposal, including the disposal site, will be decided before the start of 

dismantlement work on reactor zone's peripheral facilities, and reflected to the decommissioning plan for approval. 

<strong>The</strong> decommissioning plan for Hamaoka Nuclear Power Station Unit-1 and Unit-2 represents Japan's first 

decommissioning of a commercial light-water nuclear power plant. <strong>The</strong> priority is given to safety to steadily implement 

the decommissioning plan with transparency on Hamaoka NPS, and to acquisition of the trust from everyone concerned. 

5) 40015 – Study on Influence of Nuclear Fuel Material Management and Transfer Scenarios on 


Kazuma Mizukoshi, Nuclear Engineering, Ltd. (Japan) 

1. Summary <strong>The</strong> Japanese electric utilities are required to prepare plans to transfer its nuclear fuel material by the 

end of the decommissioning period. <strong>The</strong>re can be several scenarios regarding the management and transfer of nuclear 

fuel material. It is necessary to fully understand the characteristics of individual scenarios so that the most suitable 

method can be selected according to the conditions specific to each plant. We have examined how the nuclear fuel 

material management and transfer scenarios cause influences on the decommissioning and evaluated the characteristics 

(advantages and disadvantages) of each scenario. We expect that the result of this study will be useful for a nuclear 

power station which plans decommissioning to choose a suitable and effective scenario for the nuclear fuel material 

management and transfer. 

2. Method We collected information about methods of nuclear fuel material management and transfer and extracted 

several scenarios as shown below: Scenario A? Spent fuel (SF) is transported to a reprocessing plant after storing it in the 

same SFP as it was stored during the in-service period. Scenario B? SF is transported to an another unit of the same 

power station to transfer the management of the SF to the unit. ?<strong>The</strong> transportation between units.? Scenario C? SF is 

transported to an interim storage facility to transfer the management of SF to the interim storage facility. Scenario D: SF 

is transported to a reprocessing plant after storing it in a SFP which is isolated by replacing existing system components 

with temporarily installed components. Scenario E?SF is transported to an interim storage facility after storing it in a SFP 

which is isolated by replacing existing system components with temporarily installed components, and its management is 

transferred to the interim storage facility. 

Regarding the scenarios mentioned above, we examined the influence on decommissioning conditions (dismantling 

/ storage areas, and so on), decommissioning process, and costs. 

3. Result A study on the influence of the scenarios of nuclear fuel material management and transfer on 

decommissioning has clarified the characteristics (advantages and disadvantages) of individual scenarios and which 

scenario is suitable to specific conditions. 

4. Conclusion We could fully understand the characteristics of the scenarios of nuclear fuel material management 

and transfer. We expect the result obtained from this study will be useful for a nuclear power plant to choose a suitable 

scenario of nuclear fuel material management and transfer. 

6) 40100 – Dose Assessment for setting of EPZ in Emergency Plan for Decommissioning of Nuclear Power Plant 

Hirokazu Minato, Hitachi-GE Nuclear Energy (Japan); Takatoshi Hattori, CRIEPI (Japan); 

Toshihiko Higashi,<strong>The</strong> Kansai Electric Power Co., Inc. (Japan); Takehiro Iwata, JAPC (Japan) 

In emergency plan for a nuclear power plant, taking enough measures in EPZ boundary (Emergency Planning Zone) 

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Abstracts 

is one of priority matters, to have protection against the release of radioactive materials in accident efficiently and 

quickly to minimize environmental impacts. EPZ is set as the zone which emergency plans should be mainly prepared on 

emergency conditions. <strong>The</strong> criteria of EPZ is that dose value in the area between plant and EPZ boundary have to be less 

than 10mSv, even if very conservative release mechanisms and path are supposed. <strong>The</strong> released amount of radioactive 

material from a nuclear power plant is calculated to the accident scenario were supposed with the each phase of 

decommissioning. Moreover, the dose value is calculated as the evaluation of environmental impacts, using atmospheric 

diffusion parameters are determined by the plume concentration gaussian type distribution model at steady state, and 

annual meteorological data of the reference plants. Both of 'the spent fuel storage phase' and 'the safe maintenance and 

dismantling phase' on each of the expected accident scenario, the dose value in EPZ boundary is much less than safety 

criteria (10mSv), and there is no need to plan offsite emergency plan, such as the Sheltering and Escape for a reference 

plant. This result is agreeing with the opinion of Waste and decommission working group 2006 of Western European 

Nuclear Regulator's Association (WENRA). 

7) 40044 – Decommissioning of Ignalina NPP 

Karolis Zemkajus, Algirdas Vaidotas, Radioactive Waste Management Agency in Lithuania (Lithuania) 

<strong>The</strong> paper describes the preparation for the decommissioning and implementation of the Programme on Ignalina 

NPP decommissioning. Emphasis is given to the Final Decommissioning Plan, the whole INPP decommissioning process, 

Radiation Protection Programme, waste to be processed and disposed. Ignalina NPP was designed and constructed (Unit 

1 was commissioned in 1983, Unit 2 was commissioned in 1987) at the time when safety culture was not the issue 

addressed in a special way. <strong>The</strong> opinion of the international community was very important to Lithuania as it was striving 

for the membership of the European Union and NATO. Thus, Lithuania, respecting the opinion of the international 

community, and pursuant to the Nuclear Safety Account Grant Agreement, envisaged shutdown of Unit 1of Ignalina NPP 

by year 2005 in the National energy strategy ratified by the Seimas in 1999, considering the long-term and essential 

financial assistance conditions ensured by the EU, G-7 countries and other international institutions. <strong>The</strong> National energy 

strategy updated in year 2002 states that having in mind that the European Union countries acknowledged INPP 

decommissioning to be an exceptional financial burden for Lithuania, as well as a process which exceeds current 

financial perspectives of the country, Unit 2 will be shutdown in 2009 if sufficient financial resources supported by 

agreements with the EU institutions and other donors are in place. Regarding the preparation for the decommissioning 

and implementation of the Programme on INPP Unit 1 decommissioning during the period 2001-2004 Ignalina NPP 

developed and the Ministry of Economy approved the Final Decommissioning Plan (FDP) in 2005. This Plan covers the 

whole INPP decommissioning process (both Units, auxiliary facilities and eventually the interim spent fuel and waste 

storage facilities). Decommissioning activities and projects are planned on the basis of the herein proposed strategy. 

Generally, FDP describes principles, methods and techniques and the overall plans and schedules required to ensure 

radiological safe, environmental friendly and economic decommissioning of the INPP. In summary, the following 

information is provided in FDP: • applied regulatory framework; • adopted dismantling strategy; • plan and schedule of 

decommissioning; • radioactive and hazardous material inventories; • decommissioning activities; • decontamination 

activities; • dismantling tools and methods; • INPP waste management strategy; • safety and environmental impact 

assessment; • Radiation Protection Programme; • INPP organization performing decommissioning; • decommissioning 

costs and funding; • quality assurance; • final radiation survey and facility and site restoration. From the nuclear safety 

standpoint, the activities related to potential risk will be performed during the first 10 years following the reactor final 

shutdown (spent fuel handling, decontamination, systems modifications and isolation etc.). Works will be prepared and 

carried out by the skilled INPP staff. According the Programme on Ignalina NPP decommissioning the Schedule of 

Decommissioning: • units final shutdown: unit 1 – 31/12/2004, unit 2 – 31/12/2009; • reactors defuelling: unit 1 – 

31/12/2009, unit 2 – 28/02/2012; • pools defuelling: unit 1 – 31/08/2015, unit 2 – 12/01/2016; • start D&D activities: unit 

1 – middle 2010, unit 2 – early 2012. FDP describes volumes of unconditioned waste to be considered while preparing 

for and performing the decommissioning. In summarizing, it could be stated that INPP operation and decommissioning 

will generate approximately 5 900m3 of spent resins, perlite and sediments and about 130 000m3 of solid radioactive 

waste. <strong>The</strong> operational stored waste estimates by end 2009: • spent Fuel - 23000 assemblies; • solid Waste - 30000 m3; • 

bituminized waste - 14000 m3; • waste to be cemented (resins, perlites) - 6000 m3 (retrieval and cementation on-going 

since 2005). <strong>The</strong> Decommissioning waste estimates: • spent fuel -18000 assemblies; • equipment - 130000 tons of steel; • 

reactor - 4000 tons of steel, 4000 tons graphite; • buildings - 1000000 m3 of concrete, 200000 tons of steel. This waste 

will require processing in the new cementation facility and future solid waste management After decontamination of the 

emptied INPP facilities, final demolition of the building structures will additionally generate some 965000 m3 of 

concrete and 190000 tons of steel to be free released. 

SESSION G1 : <strong>International</strong> Collaboration 

1) 40118 – Advancing the Use of IAEA Networks in Radioactive Waste Management: Past Successes, Present 

Challenges and Future Opportunities 

Paul Degnan, John Kinker, Irena Mele, Paul J. Dinner, Horst Monken Fernandes, 

Antonio Morales, Lumir Nachmilner, Shaheed Hossain, IAEA 

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Abstracts 

Since 2001 the <strong>International</strong> Atomic Energy Agency has championed the concept and use of Networks to advance 

radioactive waste management across the globe. At the present time there are four Networks managed on behalf of 

Member States by the IAEA and a fifth one is currently being implemented. <strong>The</strong> scopes of interest covered by the 

Networks encompass near-surface and deep geological disposal, the decommissioning of nuclear facilities, the 

environmental remediation of sites contaminated with radioactive materials and the characterisation of low- and 

intermediate-level radioactive wastes. To date over 100 organisations from more than 40 Member States are involved in 

the Networks. Many of these Network participants generously donate resources, time and effort to support Network 

activities, while others with nascent or otherwise less well developed programmes are still in the process of acquiring 

experience, capabilities and know-how. Regardless of the stage of development, all Network participants share in the 

mutual benefits that arise from improved communications with sister organisations and the sharing of experience and 

knowledge. <strong>The</strong> universal Goal of the Networks is the promotion of methods and technologies that will enhance the 

safety and sustainability of radioactive waste management practices and facilities. This Goal is being achieved through 

continuous improvements in communication and knowledge sharing between Network participants and the provision of 

enhanced opportunities for training, involvement in demonstration projects and the development of novel technologies 

and methodologies. We recognise that interdisciplinary understanding and the coordination of efforts at key interfaces at 

the back-end of the fuel cycle are critical aspects for achieving the Network Goal efficiently and effectively. 

Consequently, the IAEA Networks that will be operational by the end of 2010 are themselves are being molded into an 

organic “Network of Networks” where the use of new electronic media and the possibilities presented by enhanced 

communication channels will be exploited. Here we provide an overview of the IAEA Networks in radioactive waste 

management and present a new tool that is under development, an internet-based portal for enhanced communications 

and the provision of improved training opportunities. 

2) 40287 – <strong>The</strong> activities of the OECD/NEA RWMC in the Field of HLW and SF disposal 


<strong>The</strong> OEDCD/NEA seeks to assist its member countries in developing strategies for the management of all types of 



facilities. <strong>The</strong> programme of work in the area of radioactive waste management is supervised by the Radioactive Waste 

Management Committee (RWMC) made of senior representatives from regulatory authorities, radioactive waste 

management and decommissioning organisations, policy making bodies, and research-and-development institutions from 

the NEA countries. <strong>The</strong> IAEA participates in the work of the committee, and the European Commission (EC) is a full 

member. Strong ties are maintained with national high-level advisory bodies to governments and with international 

bodies such as the <strong>International</strong> Committee on Radiation Protection. RWMC helps the national programmes through a 

broad programme of work that: fosters a shared and broad-based understanding of the state of the art and emerging 

issues; facilitates the elaboration of waste management strategies that respect societal requirements; helps to provide 

common bases to the national regulatory frameworks; enables the management of radioactive waste and materials to 

benefit from progress of scientific and technical knowledge; contributes to knowledge consolidation and transfer, e.g., 

through the organization of specialist meetings and publication of technical reports, consensus statements and short 

flyers; and helps advance the state of the art, e.g., by providing a framework for the conduct of international peer reviews. 

<strong>The</strong> latest collective statement of the RWMC dates to 2008 on “moving forward with geological disposal of radioactive 

waste”; the latest peer review was of the French Dossier 2005. A new peer review of the safety report for a spent fuel 

repository in Sweden is being organised. <strong>The</strong> RWMC holds multi-stakeholder, national workshops. <strong>The</strong> latest (2009) was 

in France, in the siting region for the national HLW repository. <strong>The</strong> RWMC manages a database of country information, 

in the form of a country reports and country profiles updated yearly; a summary of the regulatory infrastructure in NEA 

countries is also maintained. Current work areas include: promoting greater understanding of radioactive waste 

management and decommissioning disciplines; an international project on the topic of “retrievability and reversibility”; 

assisting the organisation of an “<strong>International</strong> <strong>Conference</strong> on Geological Repositories” in Japan in 2011; initiating 

dialogue with ICRP with a view to updating the ICRP guidance in the field of geological disposal; starting a project in 

the field of information and memory keeping. Specific technical areas also include optimization, treatment of the very 

long time scales, assessing the state of the art in safety assessment methods, and the operation phase of repositories. An 

overview is provided of these activities and the relevant issues. 

3) 40147 – Grimsel Test Site - Phase VI Status and Outlook 

Ingo Blechschmidt, Sven Peter Teodori, Stratis Vomvoris, Nagra (Switzerland) 

<strong>The</strong> Grimsel Test Site (GTS) (www.grimsel.com) is a generic underground research laboratory located in the 

crystalline rocks of the Aare Massif of the Swiss Alps and is owned and operated by the National Cooperative for the 

Disposal of Radioactive Waste (Nagra). <strong>The</strong> GTS is unique in that it includes a class B/C control zone in which 

repository relevant radionuclides can be used as tracers of rock processes. 2009 marked the 25th year of GTS’s operation 

and the current running Phase VI is planned until 2013. Experiments have evolved from those focused on characterising 

the structural geology, groundwater geochemistry and hydrogeology of the test site during the 80s; to radionuclide 

migration experiments in the 90s; and then more recently to assessing perturbation effects of repository implementation 

and demonstrating engineering and operational aspects of the repository system for the last 10 years. 

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Abstracts 

Currently over 25 international organisations participate in various projects at the GTS. On-going international 

partner projects are as follows: evaluation of full-scale engineered systems under simulated heat production and 

long-term natural saturation (FEBEXe); emplacement of a shotcrete low-pH plug (EC Project ESDRED); testing and 

evaluation of standard monitoring techniques (TEM/MoDeRn) including both wired and non-wired techniques; 

long-term cement studies (LCS) which aims at improving the understanding of low-pH cement interaction effects in 

water conducting features; the Colloid Formation and Migration Project (CFM) which focuses on colloid generation and 

migration from a bentonite source doped with radionuclides; and the Long-Term Diffusion (LTD) project which aims at 

in-situ verification of long term diffusion concepts for radionuclides. 

Additional experiments include techniques for determination of fracture network hydraulic and migration 

parameters and behaviour of grouting materials. New large-scale experiments to test the emplacement techniques and 

behaviour seals/plugs at 1:1 scale, under hydraulic pressure differential of up to 50 bars with parallel migration of 

repository-generated gas are in the planning stage. 

<strong>The</strong> status of the on-going experiments as well as the future plans and possibilities for new partners are summarised 

in this poster presentation. 


1) 40299 – Treatment of low level radioactive waste by plasma: a proven technology? 

Jan Deckers, Belgoprocess, NV (Belgium) 

Introduction Large amounts of actual and historical low level radioactive waste, with varying characteristics, are 

stored and generated from the operation and maintenance of nuclear power plants, the nuclear fuel cycle, research 

laboratories, pharmaceutical and medical facilities. Virtual all of these waste streams can be treated by the plasma 

technology resulting in a final product free from organics, liquids and moisture, and meets without a doubt the 

acceptance criteria for safe storage and disposal. 

Review <strong>The</strong> plasma is a highly desirable heat source. Its high temperature of up to 10,000 °C can treat the 

radioactive waste as is. <strong>The</strong> inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes 

while the organic material are vaporised into a syngas and afterwards oxidised in an afterburner. This technology is very 

suitable for historical waste containing mixtures of inorganic, organic, liquids, sludge, etc, with almost no preparation of 

the waste and minimum risks for radioactive contamination and exposure. 

Discussion In the interest of the waste producers and future generations, a high volume reduction factor (VRF) of 

the waste is desired in order to minimise the volume and overall costs of storage and waste disposal. Not only the VRF is 

of importance but also the growing requirements for improved quality of the final waste form. <strong>The</strong>refore the plasma 

technology can also be used to recondition previous conditioned waste packages that no longer meet the current 

acceptance criteria for final disposal. 

Conclusion This paper describes in detail the principles of plasma, the different waste feed systems, off gas 

treatment, operational experience and future plasma plants. 

2) 40128 – <strong>The</strong> Zwilag Plasma Facility - Five Years of Successful Operation 

Walter Heep, Zwilag Interim Storage Facility (Switzerland) 

This paper is about a treatment facility of low level radioactive wastes that operates with plasma technology.<strong>The</strong> 

first processing of low level radioctive wastes from Swiss nuclear power plants marked the successful completion of the 

commissioning of this facility in March 2004. <strong>The</strong> commercial operation of the plasma plant owned by Zwilag 

Zwischenlager Würenlingen AG (Zwilag) has also enabled this technology to be used successfully for the first time in the 

nuclear field in a way that addresses the issue of radiation protection. <strong>The</strong> plasma facility has been operating now for five 

years and was granted an unrestricted operating license in September 2009. 

3) 40293 – Safety Assessment of Disposal Container for Higher Activity Low Level Waste 

Motonori Nakagami, Seiji Komatsuki, Chubu Electric Power Co., Inc. (Japan); 

Kyosuke Fujisawa, Takashi Nishio, KOBE STEEL, LTD. (Japan); 

Thomas Quercetti, André Musolff, Karsten Müller, 

Federal Institute for Materials Research and Testing (Germany) 

As one of the studies on "yoyushindo disposal" whose concept is similar to an intermediate disposal, the 

development of a disposal container has been conducted by the Federation of Electric Power Companies of Japan. To 

assess a drop event of a waste package in which stored the radioactive wastes from nuclear power plants, the toughness 

of the disposal container was evaluated by drop tests using three specimens which have actual dimensions, drop analysis, 

fracture mechanics assessment and macroscopic tests. <strong>The</strong> three specimens for drop tests were manufactured in 

consideration of the design specifications and the manufacture operations in nuclear power plants. <strong>The</strong> lid plates of the 

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Abstracts 

specimens were welded to the body plates without pre- and post-weld heat treatment by using a remote automated 

welding machine. <strong>The</strong> drop tests showed that no penetration cracks or splash of its content occurred in the disposal 

container under conservative conditions such as the maximum weight and height in the handling. Drop analysis and the 

fracture mechanics assessment indicate that the strain induced by the drop impact did not exceed the fracture strain and 

an unstable fracture did not occur. And macroscopic tests showed that penetration cracks did not occur at 8m drop events. 

<strong>The</strong>se tests and evaluations confirmed that the disposal container had sufficient toughness. 

SESSION L5: Recycling and Clearance 

1)� 40223 – NPP Bulk Equipment Dismantling Problems and Experience 

Alexander B.Gelbutovsky, Eugeny V. Balushkin, Jury A. Epikhin, 

Alexander V. Troshev, Peter I. Cheremisin, ECOMET-S JSC (Russia) 

NPP bulk equipment dismantling problems and experience are summarized. “ECOMET-S” JSC is shown as one of 

the companies which are able to make NPPs industrial sites free from stored bulk equipment with its further utilization. 

“ECOMET-S” JSC is the Russian Federation sole specialized metallic LLW (MLLW) treatment and utilization facility. 

Company/s main objectives are waste predisposal volume reduction and treatment for the unrestricted release as a scrap. 

Leningrad NPP decommissioned main pumps and moisture separators / steam super heaters dismantling results are 

presented. Prospective fragmentation technologies (diamond and electro-erosive cutting) testing results are described. 

<strong>The</strong> electro-erosive cutting machine designed by “ECOMET-S” JSC is presented. <strong>The</strong> fragmentation technologies 

implementation plans for nuclear industry are presented too. Key words: cutting, fragmentation, NPP bulk equipment, 

recycling, utilization, main pump casing, gas cutting, melting, metal ingots, unrestricted use, separator, steam super 

heater, air plasma cutting machine, diamond cutting, electro erosive cutting, electro-erosive cutting machine. 

2)� 40073 – Reuse Of Conditionally Released Radioactive Materials From NPP Decommissioning Applied In 

Motorway Bridges Construction 

Michal Panik, Tomas Hrncir, Vladimir Necas, Slovak University of Technology in Bratislava (Slovakia) 

During the operation and especially during decommissioning of nuclear installation is produced considerable 

amount of solid materials (metals, non-metals, building structures) that can fix radioactivity in forms of contamination or 

activation. <strong>The</strong> materials present radioactive waste, part of radioactive waste may just slightly exceed limits for 

unconditional release of materials into the environment. On the other side, there is possible, after proving of defined 

safety limits fulfillment, to conditionally release radioactive waste for special purpose. In opposite case it would be 

inevitable to dispose radioactive waste in radioactive waste repository. Approaches of different countries to release of 

materials vary and the extent of this issue processing is related to each approach. Requirements set down in Slovak 

Republic legislation are given in the paper. Before the conditional release of materials there must be done consistent 

analysis of the materials impact on the inhabitants and the environment in short and long time period. <strong>The</strong> analysis 

comprises the evaluation of considered scenarios of specific utilization of conditionally released materials. This analysis 

necessarily precedes the realization of utilization. Scenarios describing utilization of radioactive waste carbon steel in the 

motorway bridge building process is stated in the paper. Radioactive steel can be utilized in many parts of the bridge. In 

the paper it is described its use as reinforcement of piles. Short time period external irradiation of workers and inhabitants 

is taken into account. Critical group (i.e. the group that gets the highest accumulated dose) of workers or inhabitants is 

chosen. Specific mass activity of released radioactive waste carbon steel is related to individual effective dose taken by 

critical group. Following legislation rules, annual effective dose taken by critical group must not overstep the limit of 10 

μSv/year. <strong>The</strong> determination of value of the specific mass activity is the target of scenarios evaluating. Evaluation of 

model scenarios can be realized with the appropriate calculation tool. In the paper VISIPLAN 3D ALARA planning tool 

was 

Session 

chosen. 

L5 - H5 - H6 

3) 40071 – Modelling of Motorway Tunnels Scenario for Utilization of Conditionally Released Radioactive 

Materials 

Tomas Hrncir, Michal Panik, Vladimir Necas, Slovak University of Technology in Bratislava (Slovakia) 

Considerable amount of solid radioactive waste is generated during the decommissioning of nuclear installations. 

Some of the materials can be released into the environment either in a direct way, if they meet the releasing limits, or 

after the application of some procedures leading to meeting the limits. If materials exceed these limits, they are treated, 

conditioned and disposed of at appropriate repository of radioactive waste. Another possible releasing way is the 

conditional release of materials, which is discussed in this paper. <strong>The</strong> basic principles of conditional release as well as 

possibilities of reusing of the conditionally released materials are described. One of these possibilities of the reusing was 

chosen and application proposal of conditional release of metal waste - steel reinforcement in the concrete, which could 

be used for construction of motorway tunnels, was created. <strong>The</strong> computer code Visiplan 4.0 3D ALARA planning tool 

software was used for the calculation of effective individual dose for personnel constructing the tunnel and for critical 

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group related to scenario. Particular models for individual scenarios of conditional release have been developed within 

the scope of this software code. <strong>The</strong> aim of the paper is to determine a level of the radioactivity of conditional released 

materials to avoid over exceeding the value of annual individual effective dose 10microSv/year established by 

international recommendations. 

4) 40117 – Estimate of Clearance Levels for Metal Materials Contaminated with Uranium 

Seiji Takeda, Hideo Kimura, JAEA (Japan) 

<strong>The</strong> Nuclear Safety Commission of Japan (NSC) published the draft report on the derivation of clearance levels for 

the solid materials contaminated with uranium (uranium-bearing wastes) in October 2009. <strong>The</strong> authors provide NSC with 

the estimated results of the clearance levels of major radionuclides, U-234, U-235 and U-238, for metal reuse scenario. 

<strong>The</strong> metal reuse scenario is categorized as two sub-scenarios on a series of recycling process of scrap metals and on reuse 

of items made from recycled metal and slag. By applying an effective dose criterion of 10?Sv/y, the activity 

concentrations are estimated from the deterministic dose calculation. <strong>The</strong> activity concentrations for U-234, U-235 and 

U-238 are calculated to be 1.5Bq/g, 1.4Bq/g and 1.8Bq/g respectively. In order to confirm the validity of the calculated 

concentrations, we estimate the uncertainties on scenario description after metal recycling and on parameter values used 

in the deterministic calculation. It is difficult to rule out the possibility that a small amount of residue of slag generated 

from recycling process is disposed of as an industrial waste. Accordingly, the authors estimate the dose for a worker 

involved with landfill disposal of residue of slag and moreover the total uranium concentration derived from the slag in 

the disposal site. <strong>The</strong> calculated dose of the worker for U-234, U-235 and U238 is about 0.05 times as low as that for the 

metal reuse scenario respectively. This result, therefore, leads to the predominance of estimating the dose on the basis of 

the metal reuse scenario. <strong>The</strong> total uranium concentration derived from the slag in the disposal site is estimated to be 

lower than the mean value of measured uranium concentration data in natural environment of Japan. This result indicates 

that the landfill disposal of residue of slag hardly brings to increase uranium concentration of natural origin. Monte 

Carlo-based uncertainty analysis was carried out in order to estimate the influence of parameter uncertainties to the result 

of deterministic calculation. <strong>The</strong> results of the uncertainty analysis, which are corresponding to the 97.5th percentile of 

minimum activity concentration estimated by applying an effective dose criterion of 100?Sv/y, are higher than those of 

deterministic calculation. <strong>The</strong> calculated activity concentrations by deterministic calculation for U-234, U235 and U-238 

were confirmed from the results of both the analysis for an additional scenario on the landfill disposal of residue of slag 

and the Monte Carlo-based uncertainty analysis. 

SESSION H5: Panel "Advances in Knowledge Management for Radioactive Waste Disposal" 


SESSION H6: Coupled Process Modeling and Natural Analogues 

1) 40306 Keynote – A Discussion of Key Issues in Coupled THM Processes in Clays, Rock Salt and Crystalline 

Rock with Bentonite Buffer 

Chin-Fu Tsang, LBNL (USA) 

Abstract Not Available 

2) 40159 – Environmental Remediation of High-Level Nuclear Waste in Geological Repository: modified 

Computer Code Creates Ultimate Benchmark in Natural systems 

Geoffrey Peter, Oregon Institute of Technology Portland Center (USA) 

Isolation of high-level nuclear waste in permanent geological repositories has been a major concern for over 30 

years due to the migration of dissolved radio nuclides reaching the water table (10,000-year compliance period) as water 

moves through the repository and the surrounding area. Repositories based on mathematical models allow for long-term 

geological phenomena and involve many approximations; however, experimental verification of long-term processes is 

impossible. Countries must determine if geological disposal is adequate for permanent storage. Many countries have 

extensively studied different aspects of safely confining the highly radioactive waste in an underground repository based 

on the unique geological composition at their selected repository location. This paper discusses two computer codes 

developed by various countries to study the coupled thermal, mechanical, and chemical process in these environments, 

and the migration of radionuclides. Further, this paper presents the results of a case study of the Magma-hydrothermal 

(MH) computer code, modified by the author, applied to nuclear waste repository analysis. <strong>The</strong> MH code verified by 

simulating natural systems thus, creating the ultimate benchmark. This approach based on processes similar to those 

expected near waste repositories currently occurring in natural systems. Keywords: High-Level Nuclear Waste, 

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Repository, Environmental Remediation, Computer Codes, Compliance Period, Benchmark. 

3) 40196 – Effect of the Residual Heat Release of the Nuclear Waste Stored in an Unsaturated Zone on 

Radionuclide Release 

Lubna K. Hamdan, John C. Walton, Arturo Woocay, University of Texas at El paso (USA) 

Over time, nuclear waste packages disposed in an unsaturated zone geological repository, such as the proposed 

repository at Yucca Mountain, are expected to fail gradually due to localized and general corrosion. As a result, water 

will have access to the nuclear waste and radionuclides will be transported to the accessible environment by ground water. 

In this paper we consider a serious failure case in which penetrations at the top and bottom of the waste package will 

allow water to flow through it (flow-through model). We introduce a new conceptual model that examines the effect of 

the residual heat release of the nuclear waste stored in an unsaturated environment on radionuclide release. This model 

predicts that the evaporation of water at the hotter sheltered areas (from condensate and seepage) inside the failed waste 

package will create a capillary pressure gradient that drives water to wick with its dissolved and suspended contents 

toward these relict areas, effectively preventing radionuclides release. We drive a dimensionless group to estimate the 

minimum length of the sheltered areas required to sequester radionuclides and prevent their release. <strong>The</strong> implications of 

this model on the performance of the proposed repository at Yucca Mountain or unsaturated zone geological repositories 

in general are explored. 

4) 40072 – Evaluation of behavior of rare earth elements based on determination of chemical state in 

groundwater in granite 

Yuhei Yamamoto, Daisuke Aosai, Takashi Mizuno, JAEA (Japan) 

Rare earth elements (REEs) in natural environment have been studied as a useful geochemical tracer and analogues for trivalent 

actinides such as Am(III) which is contained in High-Level radioactive Waste (HLW). Since Am does not occur in natural 

environment, behavior of REEs in deep groundwater gives us valuable information for research and development relating to the safety 

assessment of geological disposal of HLW. In groundwater, REEs often form complexes with aqueous and colloidal ligands according 

to their large ionic potential (ionic charge/ionic radius). <strong>The</strong>refore, behavior of REEs strongly depends on physical and chemical 

properties of these complexes. However, it is difficult to speciate complexes of REEs in groundwater mainly due to difficulties in 

direct measurement of REEs complexes with their low concentrations and alteration of groundwater condition during collection. <strong>The</strong> 

aim of this study is determination of chemical states of REEs in groundwater by combination of ultrafiltration techniques maintaining 

in-situ pressure and anaerobic conditions, speciation calculation taking care about contribution of natural organic matters, and 

finger-printing method using REE pattern of stability constants for probable complexes of REEs in groundwater. Groundwater samples 

were collected from a borehole located in the 200 m substage of the Mizunami Underground Research Laboratory (MIU), Gifu, Japan. 

Our results suggest that colloidal ligands play an important roll on the behavior of REEs in groundwater. 

5) 40022 – Natural analogue studies of bentonite reaction under hyperalkaline conditions: overview of ongoing 

work at the Zambales Ophiolite, Philippines 

Naoki Fujii, M. Yamakawa, RWMC (Japan); K. Namiki, Obayashi Corporation (Japan); 

T. Sato, Hokkaido University (Japan); N. Shikazono, Keio University (Japan); 

C. A. Arcilla, C. Pascua, University of the Philippines (Philippines); 

W Russell Alexander, Bedrock Geosciences (Switzerland) 

Bentonite is one of the most safety-critical components of the engineered barrier system for the disposal concepts 

developed for many types of radioactive waste. However, bentonite – especially the swelling clay component that 

contributes to its essential barrier functions – is unstable at high pH. To date, results from laboratory tests on bentonite 

degradation have been ambiguous as the reaction rates are so slow as to be difficult to observe in the laboratory. As such, 

a key goal in this project is to examine the reaction of natural bentonites in contact with natural hyperalkaline 

groundwaters to determine if any long-term alteration of the bentonite occurs. 

Ophiolites have been identified as sources of hyperalkaline groundwaters that can be considered natural analogues 

of the leachates produced by cementitious materials in repositories for radioactive waste. At the Zambales ophiolite in the 

Philippines, widespread active serpentinisation results in hyperalkaline groundwaters with measured pH values of up to 

11.1, falling into the range typical of low-alkali cement porewaters. <strong>The</strong>se cements are presently being developed 

worldwide to minimise the geochemical perturbations which are expected to result from the use of OPC-based concretes 

(see Kamei et al., this conference, for details). In particular, it is hoped that the lower pH of the low-alkali cement 

leachates will reduce, or even avoid entirely, the potential degradation of the bentonite buffer which is expected at the 

higher pH levels (12.5 and above) common to OPC-based concretes. During recent field campaigns at two sites in the 

Zambales ophiolite (Mangatarem and Bigbiga), samples of bentonite and the associated hyperalkaline groundwaters have 

been collected by drilling and trenching. At Mangatarem, qualitative data from a ‘fossil’ (i.e. no groundwater is currently 

present) reaction zone indicates some alteration of the bentonite to zeolite, serpentine and CSH phases. Preliminary 

reaction path modelling suggests that the zeolites could have been produced as a product of smectite reaction in the 

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hyperalkaline groundwaters. Although not included in this calculation to date, the CSH phases identified are completely 

consistent with reaction of clays with hyperalkaline groundwaters, as seen at other sites worldwide. 

At the Bigbiga site, an active hyperalkaline groundwater/bentonite reaction zone (at the base of the bentonite 

deposit) has recently been drilled and samples are currently undergoing analysis. A comparison of the data sets from both 

sites will be included in the presentation. <strong>The</strong> paper will also outline the potential future programme of research at these 

sites. 

6) 40063 – Natural Analogues of Cement: Overview of the Unique Systems in Jordan 

Gento Kamei, JAEA (Japan); 

W Russell Alexander, Bedrock Geosciences (Switzerland); Ian D. Clark, University of Ottawa (Canada); 

Paul Degnan, IAEA; Marcel Elie, Shell (UK); Hani Khoury, Elias Salameh, University of Jordan (Jordan); 

Antoni E. Milodowski, British Geological Survey (UK); Alister F. Pitty, Pitty Consulting (UK); 

John A.T. Smellie, Conterra (Sweden) 

In L/ILW repository designs, cement-based materials are expected to dominate – for example, the proposed Swiss 

repository will contain over 1.5 million tonnes of cementitious material. Models of cement evolution predict that leaching 

of the cement in the repository by groundwater will produce an initial stage of hyperalkaline (pH~13.5) leachates, 

dominated by alkali hydroxides, followed by a longer period of portlandite and CSH buffered (pH~12.5) leachates. It has 

also been predicted that the hyperalkaline porewater leached out of the near-field will interact with the repository host 

rock (and, where applicable, bentonite buffer and backfill). This could possibly lead to deterioration of those features for 

which the host rock formation and bentonite were originally chosen. 

Here, the safety assessment implications of the novel data from the Jordan Natural Analogue Study, which looked 

into interaction of natural cementitious hyperalkaline leachates on repository host rocks and clays, are presented. Several 

sites across Jordan have been studied, but the focus here will be on two particularly contrasting sites: 

• Maqarin in northern Jordan – this represents repository host rocks with advective groundwater systems. 

Hydrogeological, hydrochemical and structural data collected on the fractured rock at the site have been used to 

assess the likely implications of hyperalkaine leachate interaction on the long-term flow conditions in similar 

repository host rocks (e.g. granites, fractured sediments) and this will be discussed in detail 

• Khushaym Matruk in south-central Jordan – this represents repository host rocks with diffusive groundwater 

systems. Here, hyperalkaline leachates have diffused through an impermeable, clay-rich sediment, so providing 

information on the likely controls on leachate interaction in tight repository host rocks (e.g. claystones) 

In addition, data are provided on the nature of the secondary phases produced following interaction of the leachates 

with clays present at the sites. <strong>The</strong>se clays include mixed-layer illite/smectite and so are particularly good analogues of 

reaction of cement leachates on the bentonite buffer which is an integral part of the EBS in some L/ILW repository 

designs. This work will be contrasted with that presented by Fujii et al (this conference) which is focussed on bentonite 

reaction in leachates from low alkali cements. <strong>The</strong>se cements are under consideration for use in repositories where 

bentonite and concrete will be placed together as they produce lower pH leachates (pH 10 to 11) which are believed to be 

less aggressive to bentonite than the higher pH leachates from OPC which are discussed here. 

SESSION D4: Panel "Applying Lessons Learned from Past D&D Activities" 


SESSION R2: Environmental Remediation 

1) 40261 – Reclamation of Three In Situ Uranium Mines - Innovative Techniques 

Wallace Mays, W M Mining Company (USA) 

RECLAMATION OF THREE IN SITU URANIUM MINES IN TEXAS-INNOVATIVE APPROACHES By: 

Wallace Mays, President of IEC, Chairman and COO of Powertech Uranium, President W M Mining Company From 

1990 through 2010, Wallace Mays has been restoring the ground water and reclaiming the surface of three In Situ Leach 

Uranium Mines in south Texas; the Lamprecht, Zamzow and Pawnee In Situ Leach Mines located in Bee and Live Oak 

Counties in south Texas. <strong>The</strong>se mines were operated by Westinghouse Corporation subsidiary Wyoming Minerals and by 

Intercontinental Energy Corporation (IEC). <strong>The</strong>se were among the first In Situ Mines operated and utilized high levels of 

ammonia carbonate, which complicated the ground water reclamation project. IEC was acquired by Oren Benton, who 

declared bankruptcy in 1995 in the middle of the ground water restoration process. Westinghouse had set up an annuity 

to fund their reclamation obligations. W. Mays formed a company, Cima Energy to contract to manage the reclamation 

and to finance the funding of the working capital to be reimbursed. <strong>The</strong>se ISL mines posed very difficult ground water 

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Abstracts 

restoration and surface reclamation problems as they were developed when the industry was developing the techniques 

that later were successful. In addition, the reclamation bond was limited and required innovative techniques to restore and 

reclaim more efficiently. <strong>The</strong> paper will describe the Restoration Funding Methods and speak to issues related to 

determining bonding and how these could be improved. <strong>The</strong> paper will present a significantly enhanced ground water 

restoration procedures, describing the theory and presenting the data from both Reverse Osmosis and Ground Water 

Sweep restoration methods. Three complete ISL process facilities were dismantled, decommissioned and transported to 

licensed low level radioactive waste disposal. Innovative decontamination procedures were developed and described in 

the paper. This paper describes the restoration of pre mining ground water quality in three mines of more than 3.2 

kilometers of ground water by circulation more than 3 billion gallons of ground water through reverse osmosis, 

decontaminating, dismantling and disposal of three complete ISL Process Plants, removing, decontaminating and 

disposing of more than 50 miles of 4 inch pvc pipe from the well fields, plugging and reclaiming more than 6,600 ISL 

wells, and transporting to licensed low level radioactive waste disposal sites more than 650 trucks of low level waste. 

Surface reclamation procedures and problems will be presented. 

2) 40005 – Environmental remediation Activities at the Ningyo-toge Uranium Mine, Japan 

Hiroshi Saito, Tomihiro Taki, JAEA (Japan) 

Ningyo-toge Uranium Mine is located at and around the boundary of Okayama and Tottori Prefectures, western part 

of Japan. In the Ningyo-toge Mine, exploration activities had been carried out in 1950’s and 60’s after the outcrop was 

discovered in 1955. Mining activities using galleries and an open-pit had been carried out for sending uranium ores to the 

conversion and the enrichment plants to 1987 when the mining activities were terminated to begin the environmental 

remediation activities. 

<strong>The</strong>re are many facilities subject to the environmental remediation, including a mill tailings pond, a former open-pit 

mine and waste rock yards. <strong>The</strong> main purposes of the environmental remediation common to these facilities, are to take 

measures to reduce the radiation exposure from the exposure pathways to humans in future, and to prevent the occurrence 

of relevant environmental contamination. 

So far, a great number of data have been acquired and technical methods have been examined for the future 

remediation. And using the above-mentioned data, JAEA has been conducting the remediation activities at the related 

facilities. Among them, the mill tailings pond, operated since 1965 with the approved volume about 40,000m3, has 

deposited mining waste and impounded mine water as a buffer reservoir before it is transferred to the water treatment 

facility. It is located at the upstream of the water-source river, and therefore, its presence is a cause for worry to the local 

residents. Also, social impact is thought to be extensive in case of an outflow incident of mill tailings, like dam failure by 

the earthquake. Thus the highest priority has been put to the pond. 

JAEA has planned to conduct the remediation and close the pond in coming couple of years. Some activities have 

already begun, and the results have been produced steadily. According to the current plan, the pond will be covered by 

the multi-layered capping following dewatering and reshaping of mill tailings. <strong>The</strong> capping is composed of “radon 

barrier” for lowering radon-gas dissipation and dose rate, and “low-permeable protective layer” for protecting the radon 

barrier and reducing the amount of permeated rainwater. Natural material, including bentonite and sand, is planned for 

use to alleviate the future maintenance. Currently, designing is underway for the upstream half of the pond. Data will be 

accumulated after capping to verify its effectiveness, and if proved effective, it will be utilized for the capping of the 

downstream half of the pond. 

3) 40092 – Radon impact at a remediated uranium mine site in Japan 

Yuu Ishimori, JAEA (Japan) 

This paper mainly illustrates the radon impact of the closed uranium mine site remediated in 2007. <strong>The</strong> site 

remediated is the waste rock site located on the steep slope of a hill about1.5 km upstream from a residential area along a 

main ravine. Major remedial action was to cover these waste rock yards with weathering granite soil. <strong>The</strong> radon flux 

density after remediation was intended to be 0.1 Bqm-2s-1 in consideration with the natural background level around 

Ningyo-toge because there is no value of radon flux density regulated in Japan. Our action decreased the radon 

concentration in the site to natural background level, approximately from 10 to 40 Bqm-3, although relatively high 

concentration in excess of 100 Bqm-3 was observed before remediation. On the other hand, our action did not decrease 

the radon concentrations around the site in general. This fact proved that the limited source such as waste rocks affected 

the radon concentrations at neighboring area only. <strong>The</strong> similar tendencies were also observed in other environmental data 

such as radon progeny concentrations. In conclusion, these findings proved that our remedial action was successful 

against radon. This fact will lead to more reasonable action plans for other closed mine sites. 

4) 40243 – Phosphate based remediation techniques: interaction of phosphate with uranium-bound calcite 

Chase Bovaird, Dawn Wellman, PNNL (USA) 

Despite several decades of studies, effective uranium cleanup strategies remain elusive for contamination in deep 

subsurface settings that prevail in a number of Department of Energy sites in the western USA. Numerous strategies have 

been proposed, including iron barriers, soluble reductive agents, and microbial stabilization via reduction and 

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precipitation, but have limited applicability for deep subsurface remediation in an oxidative environment. In-situ 

phosphate based remediation techniques can potentially delay the precipitation of phosphate phases for controlled in situ 

precipitation of stabile phosphate phases to control the long-term fate of uranium. <strong>The</strong> basic principles underlying 

phosphate stabilization is that aqueous phosphate (PO43-), whether injected as an aqueous solution or solubilized from a 

source reacts with heavy metals to form insoluble metal-phosphate minerals. <strong>The</strong> sorption of uranyl species onto minerals 

is dependent on the nature and availability of binding sites, solution composition and pH, and aqueous complexation. 

Uranium can be sequestered either through ion exchange or surface complexation, and the rates of release of uranium are 

dependent on the sorption mechanism. <strong>The</strong> Hanford Site in southeastern Washington State is a former nuclear defense 

production facility. Uranium has been identified as a contaminant of concern for groundwater and the deep vadose zone. 

<strong>The</strong> vadose zone is comprised of highly alkaline, calcareous sediment. EXAFS analyses at shallow depths suggest that 

uranium-rich calcite is one of the major controlling phases. Calcite can also serve as a source of Ca2+ and CO32- ions to 

form mobile, aqueous, uranyl-carbonate species [Ca2UO2(CO3)3] under circumneutral to alkaline conditions. Detailed 

understanding of the rate and mechanism of the interaction between phosphate and uranium-rich calcite will allow a more 

effective design of aqueous phosphate-based infiltration strategies to minimize the mobilization of uranium during 

remediation. <strong>The</strong> objective of this investigation was to evaluate the interaction of phosphate species with uranium-rich 

calcite to determine the effects of geochemical conditions on the partitioning of phosphate and its degradation products 

with uranium-rich calcite, quantify the release of uranium from uranium-rich calcite based on the identity and 

concentration of aqueous phosphate species, and quantify the rate and mechanism of uranium immobilization based on 

the identity and concentration of aqueous phosphate species. <strong>The</strong> information obtained from this line of inquiry is 

essential to effectively develop phosphate-based remediation strategies for uranium in calcareous environments. 

5) 40220 – Remediation of Old Environmental Liabilities in <strong>The</strong> Nuclear Research Institute Rez plc 

Karel Svoboda, Josef Podlaha, Nuclear Research Institute Rez plc (Czech Republic) 

<strong>The</strong> Nuclear Research Institute Rez plc (NRI) after 55 years of activities in the nuclear field produced some 

environmental liabilities that shall be remedied. <strong>The</strong>re are three areas of remediation: (1) decommissioning of old 

obsolete facilities (e.g. decay tanks, RAW treatment technology, special sewage system), (2) processing of RAW from 

operation and dismantling of nuclear facilities, and (3) elimination of spent fuel from research nuclear reactors operated 

by the NRI. <strong>The</strong> goal is to remedy the environmental liabilities and eliminate the potential negative impact on the 

environment. Remediation of the environmental liabilities started in 2003 and will be finished in 2014. <strong>The</strong> character of 

the environmental liabilities is very specific and requires special remediation procedures. Special technologies are being 

developed with assistance of external subcontractors. <strong>The</strong> NRI has gained many experiences in the field of RAW 

management and decommissioning of nuclear facilities and will use its facilities, experienced staff and all relevant data 

needed for the successful realization of the remediation. <strong>The</strong> most significant items of environmental liabilities are 

described in the paper together with information about the history, the current state, the progress, and the future activities 

in the field of remediation of environmental liabilities in the NRI. 

SESSION G2: IAEA Topical for Disused Sealed Radioactive Sources (DSRS) 

1) 40028 – <strong>International</strong> initiatives addressing the safety and security of Disused Sealed Radioactive Sources 

(DSRS) 


High activity radioactive sources provide great benefit to humanity through their utilization in agriculture, industry, 

medicine, research and education, and the vast majority is used in well-controlled environments. None-the-less, control 

has been lost over a small fraction of those sources resulting in accidents of which some had serious – even fatal – 

consequences. Indeed, accidents and incidents involving radioactive sources indicate that the existing regime for the 

control of sources needs improvement. Additionally, today’s global security environment requires more determined 

efforts to properly control radioactive sources. Consequently, the current regimes must be strengthened in order to ensure 

control over sources that are outside of regulatory control (orphan sources), as well as for sources that are vulnerable to 

loss, misuse, theft, or malicious use. Besides improving the existing situation, appropriate norms and standards at the 

national and international levels must continue to be developed to ensure the long-term sustainability of control over 

radioactive sources. In order to improve the existing situation, concerted national and international efforts are needed and, 

to some degree, are being implemented to strengthen the safety and security of sources in use, as well as to improve the 

control of disused sources located at numerous facilities throughout the world. More efforts must also be made to identify, 

recover, and bring into control orphan sources. <strong>The</strong> IAEA works closely with Member States to improve the safety and 

security of radioactive sources worldwide. Besides the IAEA Technical Assistance Programme and Technical 

Cooperation Fund, donor States provide significant financial contributions to the Nuclear Security Fund and/or direct 

technical support to other States to recover, condition and transfer disused sources into safe and secure storage facilities 

and to upgrade the physical protection of sources that are in use. Under the USA-Russian Federation-IAEA (“Tripartite”) 

Initiative, for example, disused sources of a total activity of 2120 TBq (57251 Ci) were recovered and transported into 

safe and secure storage facilities in six countries of the former Soviet Union. Additionally, physical protection upgrades 

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were performed in thirteen former Soviet Union republics at facilities using or storing high activity radioactive sources. 

Canada has also provided funding for projects related to the safety and security of radioactive sources in the same region. 

Additionally, the EU and other countries are making regular and significant contributions to the IAEA for projects aimed 

at upgrading the safety and security of radioactive sources in South-Eastern Europe, the Middle East, Asia and Africa. 

Depending on the status of the radioactive source (in use, disused, or orphan) and the actual technical, safety and security 

situation, several options exist to ensure the source is properly brought or maintained under control. This paper will 

describe those options and the systematic approach followed by the IAEA in deciding on the most appropriate actions to 

take for the high activity sources that need to be recovered or removed from the countries under that request assistance. 

2) 40303 – Current situation and Management Plan of Radioactive Sources in Japan 

Hirokuni Ito, Tadashi Ishii, Tomokazu Ueta, Takao Nakaya, Kenya Suyama, MEXT (Japan) 

1, Status of Radioactive Material Distribution in Japan Almost all radioactive products in Japan are directly 

imported or produced by using imported radioactive materials. Approximately 99% (in terms of radioactivity; as of 2009) 

of sealed radioactive sources distributed in Japan are imported and supplied through Japan Radio Isotope Association 

(JRIA). 

2, Disused Sealed Radioactive Sources (DSRS) and Orphan sources Substantially, in Japan the users of the sealed 

radioactive sources will return to manufacturers or distributors. JRIA collects the DSRS based on the sales contract 

between JRIA and the purchasers of the sealed radioactive sources. <strong>The</strong> collected DSRS are tested in JRIA’s facility to 

check the surface contamination and the radioactivity for the transport safety. JRIA also receives orphan radioactive 

sources in Japan in accordance with MEXT’s administrative instruction if they are found. 

JRIA returns collected DSRS, except for the products using very short half life nuclides, to the radioactive source 

manufacturers in accordance with international guidelines and the import agreement between JRIA and the source 

manufacturers. 

JRIA keeps the collected DSRS in their own storage facility in cases where they can not return them to the 

radioactive source manufacturers. However, generally speaking, such DSRS kept temporary in the storage facility of 

JRIA have been returned successively to the manufacturer, through the discussion among relevant parties if necessary. In 

Japan, by the system that JRIA receives disused sealed sources based on the sales contract mentioned above, orphan 

sources are scarcely found. Nevertheless, in such case, they will be picked up and kept safely in storage facility of JRIA. 

3, Long term Management Plan of DSRS in Japan <strong>The</strong>re is no disposal facility for DSRS in Japan. But as we 

explained above, it is not the serious issue which should be resolved soon. This is because i) almost all of DSRS are 

returned to source manufacturers, ii) JRIA has still enough storage capacity, iii) and the number of DSRS stored in JRIA 

is settled. 

We understand the potential necessity of the disposal facility of DSRS depending on the number of stored DSRS. 

But considering current situation, it is not required to construct the additional storage facility or the disposal facility for 

DSRS in Japan. Of course, this view depends on the increase rate of stored DSRS and the capacity of the storage facility 

in future. We would cooperate with the international society and keep attention to the status of DSRS in Japan for the 

safety regulation of the radioactive source considering security aspects. 

3) 40060 – <strong>The</strong> Deployment of the Mobile Hot Cell to Condition High Activity Disused Sealed Radioactive 

Sources (DSRS) for Long Term Storage or Removal 

Gerhardus R. Liebenberg, South African Nuclear Energy Corporation (Necsa) (South Africa) 

<strong>The</strong> <strong>International</strong> Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. 

National Nuclear Security Agency (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, 

evaluation, and testing of a mobile hot cell intended to address the problem of high activity disused sealed radioactive 

sources (DSRS) in obsolete irradiation devices such as teletherapy heads and dry irradiators. 

Operations to condition high activity DSRS using the mobile hot cell has successfully been undertaken in various 

countries since April 2009. <strong>The</strong> project was initially targeting the African continent but is now also expanding to other 

parts of the world such as Latin America and Asia. <strong>The</strong> mobile hot cell allows for source removal, characterization, 

consolidation, repackaging in stainless steel capsules as special form, and secure storage of high risk DSRS in modern 

long term storage shields at single sites in each IAEA Member State. 

<strong>The</strong> mobile hot cell and related equipment is transported in two shipping containers to a specific country where the 

following process takes place: - Assembly of hot cell - Removal of high activity DSRS from working shields, 

encapsulation into a stainless steel capsule to obtain special form status and placement into a long term storage shield - 

Conditioning of any other spent sources the country may require. - Dismantling of the hot cell - Shipping equipment out 

of country. 

This presentation will discuss the design of the mobile hot cell as well as the deployment of the unit for 

manipulation of high activity DSRS in various countries worldwide. As a result of this project, excess high activity 

DSRS could be managed safely and securely and possibly be more easily repatriated to their country of origin for 

appropriate final disposition. 

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4) 40266 – Problems with Packaged Sources in Foreign Countries 

James Matzke, John Zarling, Cristy Abeyta, Joseph A. Tompkins, LANL (USA) 

<strong>The</strong> Global Threat Reduction Initiative’s (GTRI) Off-Site Source Recovery Project (OSRP), which is administered 

by the Los Alamos National Laboratory (LANL), removes excess, unwanted, abandoned, or orphan radioactive sealed 

sources that pose a potential risk to health, safety, and national security. In total, GTRI/OSRP has been able to recover 

more than 19,000 excess and unwanted sealed sources from over 750 sites. In addition to transuranic sources, the 

GTRI/OSRP mission now includes recovery of beta/gamma emitting sources, which are of concern to both the U.S. 

government and the <strong>International</strong> Atomic Energy Agency (IAEA). This paper provides a synopsis of cooperative efforts 

in foreign countries to remove sealed sources by discussing three topical areas: 1) <strong>The</strong> Regional Partnership with the 

<strong>International</strong> Atomic Energy Agency; 2) Challenges in repatriating sealed sources; and 3) Options for repatriating sealed 

sources. 

5) 40085 – <strong>The</strong> IAEA's approach to the security of radioactive material 


Over the past decade, the threat has increased of terrorism and other malevolent acts by terrorist groups and other 

malicious non-State actors, involving the potential use of radioactive materials. This has led to an international effort to 

build a nuclear security framework and regime, both for prevention and consequence management. Legally binding and 

non-binding international instruments have been established that form the international framework for an effective 

nuclear security regime. Adherence to and implementation of these instruments is vital for effective nuclear security. 

IAEA implements a comprehensive programme to assist States in strengthening their nuclear security. <strong>The</strong> third Nuclear 

Security Plan covers the period 2010–2013. Through the implementation of these plans, IAEA conducts advisory 

services and provides technical advice, support and training. It also addresses the longer-term effort of development of 

nuclear security guidance and it facilitates outreach and information exchange through databases, conferences, 

workshops and fellowships. Nuclear security issues relating to the prevention and detection of, and response to, theft, 

sabotage, unauthorized access and illegal transfer or other malicious acts involving nuclear material and other radioactive 

substances and their associated facilities are addressed in the IAEA Nuclear Security Series of publications. <strong>The</strong>se 

publications are consistent with, and complement, international nuclear security instruments, such as the Code of 

Conduct on the Safety and Security of Radioactive Sources. Nuclear security missions, evaluations and technical visits 

continue to be the Agency’s main tool for helping States to assess their nuclear security needs, and provide a basis for 

formulating plans of action for improving nuclear security. <strong>The</strong> needs identified by such missions can be subsequently 

addressed by the State alone, in conjunction with Agency support, or with the assistance of a bilateral partner. Annually, 

the IAEA conducts more than 60 training events for Member States as well as for Non-Member States. <strong>The</strong>se are based 

on findings and insights resulting from the various advisory missions and organized in response to the requests 

formulated by the States themselves. Providing urgently needed technical upgrades and equipment has been a foundation 

for IAEA assistance to States in enhancing the security of radioactive material since the establishment of the Nuclear 

Security Programme. <strong>The</strong> equipment needed was provided to States as follow-up to assessment missions and the training 

needed to operate the equipment was arranged in separate events. 

6) 40058 – Radioactive Waste Management in Lebanon 

Munzna Assi, Lebanese Atomic Energy Commission (Lebanon) 

<strong>The</strong> disused sealed radioactive sources including orphan sources in Lebanon, along with the growing industry of 

sealed radioactive sources in medical, industrial and research fields have posed a serious problem for authorities as well 

as users due to the lack of a national store for disused radioactive sources. Assistance form <strong>International</strong> Atomic Energy 

Agency (IAEA) was requested to condition and store disused radium needles and tubes present at two facilities. <strong>The</strong> 

mission took place on July 25, 2001 and was organized by the (IAEA) in cooperation with the Lebanese Atomic Energy 

Commission (LAEC). Other disused radioactive sources were kept in the facilities till a safer and securer solution is 

provided; however orphan sources, found mainly during export control, were brought and stored temporarily in LAEC. 

<strong>The</strong> necessity of a safe and secure store became a must. Prior to October 2005, there was no clear legal basis for 

establishing such store for disused radioactive sources, until the ministerial decree no 15512 dated October 19, 2005 

(related to the implementation of decree-law no 105/83) was issued which clearly stated that “<strong>The</strong> LAEC (Lebanese 

Atomic Energy Commission) shall, in cooperation with the Ministry of Public Health, establish a practical mechanism 

for safe disposal of radioactive waste”. Following this, the work on inventory of disused sealed sources along with 

collecting orphan sources and placing them temporarily in LAEC was legally supported. Moreover, several missions 

were planned to repatriate category I and II sources, one of which was completed specifically in August 2009; other 

missions are being worked on. In 2008, a national technical cooperation project with the <strong>International</strong> Atomic Energy 

Agency, IAEA, was launched. Under the reference TC number LEB3002, the project was entitled “"Assistance in the 

establishment of a safe temporary national storage at the Lebanese Atomic Energy Commission for orphan sources and 

radioactive waste" which cycle is 2009-2011. Under this project, a national store for radioactive sources in the third 

basement of LAEC is being established. <strong>The</strong> area is being reconstructed currently and will be equipped when ready under 

LEB3002 project. Along with this, a system for sealed disused sources management has been prepared, part of which is 

applied now and the rest will be applied upon the establishment of the store. This paper will cover the inventory 

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Abstracts 

collection process, the study for the establishment of this store, the present and prospective waste management system, 

and the the waste acceptance criteria. 

7) 40029 – <strong>The</strong> ultimate solution – disposal of Disused Sealed Radioactive Sources (DSRS) 


For countries with no access to existing or planned geological disposal facilities for radioactive wastes, the only 

options for managing high activity or long-lived disused radioactive sources are to store them indefinitely, return them to 

the supplier or find an alternative method of disposal. Disused sealed radioactive sources (DSRS) pose an unacceptable 

radiological and security risk if not properly managed. Out of control sources have already led to many high-profile 

incidents or accidents. Countries without solutions in place need to consider the future management of DSRSs urgently. 

In the frame of a regional Technical Cooperation (TC) project, a number of countries in the African region have come 

together under the auspices of the IAEA to explore the option of a Borehole Disposal Concept (BDC) for their small 

inventories of DSRSs. 

Disposal in boreholes is intended to be simple and effective, meeting the same high standards of long-term 

radiological safety as any other type of radioactive waste disposal. It is believed that the BDC can be readily deployed 

with simple, cost-effective technologies. <strong>The</strong>se are appropriate both to the relatively small amounts and activities of the 

wastes and the resources that can realistically be found in African countries. <strong>The</strong> South African Nuclear Energy 

Corporation Ltd (Necsa) has carried out project development and demonstration activities since 1996. <strong>The</strong> project looked 

into the technical feasibility, safety and economic viability of BDC under the social, economic, environmental and 

infrastructural conditions currently prevalent in Africa. Conceptually, the disposal concept comprises a borehole 

(diameter ranging from 150 to 260 mm) drilled down to a depth ranging from 30 to 100 metres. <strong>The</strong> depth will be 

dependent on the site-specific safety assessment. <strong>The</strong> borehole will have a casing with a plug at the bottom. Grouting will 

then be applied to seal the annulus and all fractures and crevices outside of the casing. <strong>The</strong> spacing of waste packages is 

~1 metre. <strong>The</strong> space between packages is backfilled with a suitable material such as cement or concrete grout. In the 

generic design of the Borehole Disposal Concept, a 100-meter deep borehole is to be filled with 50 packages up to the 

depth of 50 metres. <strong>The</strong> rest will be backfilled with concrete to act as a borehole plug. 

<strong>The</strong> project was completed in 2004 and subject to an IAEA peer review by an independent group of international 

experts in April 2005. One of the main outcomes of the international peer review was the positive statement by the expert 

team that the BDC developed by Necsa had been demonstrated to be a safe, economic, practical and permanent means of 

disposing of disused radioactive sealed sources. 

Implementation is near at hand with work being done in Ghana with support from the IAEA. Here the site selection 

is complete and studies are being carried out to test the site parameters for inclusion into the safety assessment. 

SESSION L6: Waste Treatment 

1) 40055 – Drying System For Radioactivated Metal Waste From Nuclear Power Station 

Nobuhito Ogaki, Yasushi Ooishi, Hironori Takabayashi, Masamichi Obata, Taichi Horimoto, 

Toshiba Corporation (Japan) 

High dose rate metal waste from core internals, BWR channel box or control rod is stored in fuel pool or site bunker 

pool. Waste form for final disposal of these high dose rate metal waste should elinimate water to prevent hydrogen gas 

which is caused by radialysis of water. Toshiba's newly developed drying system enables short drying time and easy 

maintenance. Toshiba will provide the total system to fabricate the waste form for high dose rate metal waste. 

2) 40186 – Macroporous Catalysts for Hydrothermal Oxidation of Metallorganic Complexes at Liquid 

Radioactive Waste Treatment 

Valentin Avramenko, Dmitry Marinin, Vitaly Mayorov, Alexander Mironenko, Marina Palamarchuk, 

Valentin Sergienko, Institute of Chemistry FEDRAS (Russia) 

One of the main problems of liquid radioactive waste (LRW) management is concerned with treatment of 

decontamination waters containing organic ligands. <strong>The</strong> organic ligands like oxalic, citric and ethylenediaminetetraacetic 

acids form stable complexes with radionuclides which puts restrictions on application of many technologies of LRW 

management. One of the ways of destruction of metallorganic complexes consists in using the catalytic oxidation. 

However, the heterophase catalytic oxidation is rather problematic due to formation of metal oxides on the catalyst 

surface and calmatation of meso- and micropores. A possible solution of the above problem can be found in synthesis of 

macroporous catalysts for oxidation having a regular macroporous structure. <strong>The</strong> present paper describes the template 

synthesis of macroporous metalloxide catalysts performed with using siloxane-acrylate microemulsions as templates. <strong>The</strong> 

method for impregnation of precious metals (PM) particles into the template, which enables one to produce PM 

nanoparticles of a specific size and immobilize them in the porous structure of the synthesized metalloxide catalysts, is 

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Abstracts 

presented. A possible mechanism of the synthesis of macroporous catalysts is suggested and the comparison of the 

electronic and photon-correlation spectroscopy results obtained at different stages of catalysts synthesis was conducted. 

3) 40157 – Carbonaceous radioactive waste treatment 

Gerard Laurent, EDF/CIDEN (France) 

In the carbonaceous waste, the aim is to isolate and confine for storage purposes the volatile radionuclides such as 

tritium (3H), chlorine-36 (36Cl) and, notably, carbon isotopes, in particular radioactive isotope 14C (referred to hereafter 

as "carbon-14"). 

For this purpose, the invention includes two types of treatment of this carbonaceous waste: - a first type of treatment 

to obtain a carbon oxide, such as carbon monoxide and/or carbon dioxide, the carbon element of which is the isotope 

carbon-14, and - a second type of treatment of this carbon oxide to obtain a solid precipitate by reaction with a chosen 

element such as, for example, calcium. 

This second type of treatment, referred to as "carbonation", consists, for example, in bubbling the carbon oxide 

through a solution containing quick lime (when the chosen element is calcium). <strong>The</strong> solid precipitate obtained (typically 

CaCO3 calcite, the carbon element of which is the isotope 14C) may then be confined for long-term storage in bulk in 

containers stored above-ground or buried under a layer of earth of a certain thickness, beneath a hill, for example. A 

variant shall be noted here, consisting in making the carbon oxide react with an element other than calcium, such as 

magnesium (or other metals), to obtain magnesite MgCO3. Essentially, this second type of treatment generally aims to 

obtain an insoluble solid precipitate, made up of carbonates and/or salts comprising the carbon element. 

4) 40163 – Impermeable graphite: A new development for the waste management of irradiated graphite 

Johannes Fachinger, Karl-Heinz Grosse, Furnances Nuclear Applications Grenoble (Germany); 

Richard Seemann, Milan Hrovat ALD (Germany) 

Graphite is a geological stable material proven by its natural occurrence. However its porous structure anticipates 

the use of graphite as long term stable waste matrix for final disposal. A slow corrosion in aquatic phases can be induced 

by high irradiation doses in the range of 10E-5 to 10E-7 g/m²d. <strong>The</strong> porous structure is related to large surface areas and 

therefore the radiation induced corrosions process cannot be neglected for final disposal. Furthermore aqueous phases 

will penetrate into the pore system and radionuclides adsorbed on the surface will be dissolved. All this problems can be 

solved with a graphite material with a closed pore system. A graphite composite material with an inorganic binder has 

been developed with a density > 99,9 % of theoretical density and therefore a negligible porosity. A first draft calculation 

predicts that the life time of such a material will be at least 2 orders of magnitude higher than porous graphite. This 

material represents a long term stable leaching resistant matrix for the embedding of i-graphite. Granulated i-graphite will 

be mixed with natural graphite and inorganic binder and pressed as block. Other radioactive wastes can be embedded in 

this matrix additionally, e.g. coated particles of spent fuel from HTR reactors. First investigations proofed the expected 

pore free structure and good mechanical strength properties. <strong>The</strong> paper further will present a selection of potential 

additonal applications of the new material as embedding material for of different nuclear waste streams like noble metals 

from reprocessing or iodine. 

5) 40165 – THOR® Steam Reforming Technology for the Treatment of Ion Exchange Resins and More 

Complex Wastes such as Fuel Reprocessing Wastes 

J. Brad Mason, Corey Myers, Studsvik, Inc. (USA) 

<strong>The</strong> THOR fluid bed steam reforming process has been successfully operated for more than 10 years in the USA for 

the treatment of low- and intermediate-level radioactive wastes generated by commercial nuclear power plants. <strong>The</strong> 

principle waste stream that has been treated is primarily ion exchange resins but various liquids, sludges and solid 

organic wastes have also been treated. <strong>The</strong> primary advantages of the THOR process include: (a) maximum volume 

reduction on the order of 5:1 to 10:1 depending on the waste type and waste characteristics; (b) environmentally 

compliant off-gas emissions, (c) reliable conversion of wastes into mineralized products that are durable and 

leach-resistant, and (d) no liquid effluents. 

Over the past five years, the THOR process has been adapted for the treatment of more complex wastes including 

historic defense wastes, reprocessing wastes, and other wastes associated with the fuel cycle. As part of U. S. Department 

of Energy (DOE) environmental remediation activities, the THOR dual bed steam reforming process was successfully 

demonstrated to process: Idaho National Laboratory (INL) Sodium-Bearing Waste (SBW); Savannah River Tank 48 

High Level Waste (HLW); and Hanford Low Activity Waste (LAW) and Hanford Waste Treatment Plant Secondary 

Waste (WTP SW) liquid slurry simulants. <strong>The</strong> THOR process has been demonstrated in pilot plant operations to 

successfully process various simulated liquid, radioactive, nitrate-containing wastes into environmentally safe, 

leach-resistant solid mineral products. <strong>The</strong> solid products incorporate normally soluble ions, such as sodium, potassium, 

cesium, technetium, and sulfate, chloride and fluoride salts into an alkali alumino-silicate mineral matrix that inhibits the 

leaching of those ions into the environment. <strong>The</strong> solid mineral products produced by the THOR process exhibit durability 

and leach resistance characteristics that are superior to borosilicate waste glasses. As a result of this work, a full-scale 

THOR process facility is currently under construction at the DOE’s Idaho site for the treatment of SBW and a full-scale 

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Abstracts 

facility is in the final design stage for the DOE’s Savannah River Site for the treatment of Tank 48 Waste. 

Recent work has focused on the development of new monolithic waste formulations, the extension of the THOR 

process to new waste streams, and the development of modular THOR processes for niche waste treatment applications. 

This paper will provide an overview of current THOR projects and summarize the processes and outcomes of the 

regulatory and safety reviews that have been necessary for the THOR process to gain acceptance in the USA. 

6) 40257 – Phase Behavior and Reverse Micelle Formation in Supercritical CO2 with DTAB and F-pentanol 

for Decontamination of Radioactive Wastes 

Kensuke Kurahashi, Osamu Tomioka, Yoshihiro Meguro, JAEA (Japan) 

Decontamination of radioactive wastes is useful for the volume reduction and re-categorization of them. However, a 

large amount of secondary wastes are often generated from a wet decontamination using traditional solvents. <strong>The</strong>refore, 

the development of decontamination method decreasing the secondary wastes is required. Supercritical CO2 (scCO2), 

which is gaseous matter at ordinary conditions, has the potential to minimize the amount of solvent wastes. However, 

neat scCO2 is not suitable for the separation of polar materials, such as metal compounds, because non-polar scCO2 

cannot dissolve them. A reverse micelle is a stable aggregate of amphiphilic surfactants surrounding a water pool in a 

non-polar solvent. Polar materials can be dissolved in the water pool and dispersed in the nonpolar solvent. <strong>The</strong>refore, the 

formation of reverse micelle in scCO2 could increase the availability of scCO2 as a solvent for separation of 

radionuclides from radioactive wastes. In the present study, we investigated the reverse micelle formation in scCO2 with 

dodeceytrimethylammonium bromide (DTAB) and 2,2,3,3,4,4,5,5-octafluoro-1-pentanol (Fpentanol). After putting a 

mixture of DTAB, F-pentanol and water in a high pressure view cell of 54 cm3 at 45 °C, CO2 was introduced into the 

cell slowly. When the mixture was dissolved completely in the scCO2, the pressure was measured as the cloud point 

pressure (CPP). When 1.08 mmol DTAB were employed, 13.5 mmol H2O could be dissolved in scCO2 with 6.3 vol% 

F-pentanol at 20.9 MPa. Formation of reverse micelles was confirmed by dissolving an aqueous solution of methyl 

orange into the scCO2 with DTAB and water. <strong>The</strong> CPP values were determined for various volume of water. <strong>The</strong> CPP 

values increased slowly with an increase of water concentration in the cell less than 0.35 mol/dm3. <strong>The</strong> increase of water 

concentration led the growth of reverse micelle size, and then the CPP value would rise. On the other hand, the CPP did 

not simply increase with the water concentration more than 0.35 mol/dm3 and a sharp and concave peak appeared in the 

relationship between the CPP value and the water concentration. This suggests that the state of water in the scCO2 

changed before and after the water concentration of 0.35 mol/dm3. <strong>The</strong> water concentration giving the CPP peak 

depended on the F-pentanol concentration but was independent of the DTAB concentration. From this fact, it is expected 

that the interaction between F-pentanol and water affects the state of water in the scCO2. 

SESSION H7: Performance Assessment Modeling and Parameters 

1) 40305 Keynote – Development of a Realistic Repository Performance Assessment Method 


Abstract Not Available 

2) 40204 – Integrated model for the near field of a repository in granite host-rock. Probabilistic approach 

Lara Duro, Alba Valls, Olga Riba, Jordi Bruno, Amphos XXI Consulting S.L.(Spain); 

Aurora Martinez-Esparza, ENRESA (Spain) 

<strong>The</strong> application of probabilistic approaches to the performance of underground repositories for long-lived 

radioactive waste has received special attention in the last years. Numerous exercises have been developed in order to 

elicit the Probability Distribution Functions (PDFs) of the several parameters needed for these developments. Several 

integrated models allow the implementation of PDFs in the long-term simulations needed for Performance Assessment. 

In this work we present how the deterministic compartmental model for a repository of high level nuclear waste (HLNW) 

located in a crystalline host-rock has been modified to include PDFs for some of the parameters. <strong>The</strong> implementation of 

probabilistic approaches gives also information on the most influencing parameter on the migration of the different 

radionuclides from a deep repository concept. 

3)� 40017 – Spatial Variability and Parametric Uncertainty in Performance Assessment Models 

Osvaldo Pensado, James Mancillas, Scott Painter, Southwest Research Institute (USA); 

Yasuo Tomishima, AIST (Japan) 

<strong>The</strong> problem of defining an appropriate treatment of distribution functions (which could represent spatial variability 

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Abstracts 

or parametric uncertainty) is examined based on a generic performance assessment model for a high-level waste 

repository. <strong>The</strong> generic model incorporated source term models available in GoldSim®, the TDRW code for contaminant 

transport in sparse fracture networks with a complex fracture-matrix interaction process, and a biosphere dose model 

known as BDOSE(TM). Using the GoldSim framework, several Monte Carlo sampling approaches and transport 

conceptualizations were evaluated to explore the effect of various treatments of spatial variability and parametric 

uncertainty on dose estimates. Results from a model employing a representative source and ensemble-averaged pathway 

properties were compared to results from a model allowing for stochastic variation of transport properties along 

streamline segments (i.e., explicit representation of spatial variability within a Monte Carlo realization). We concluded 

that the sampling approach and the definition of an ensemble representative do influence consequence estimates. In the 

examples analyzed in this paper, approaches considering limited variability of a transport resistance parameter along a 

streamline increased the frequency of fast pathways resulting in relatively high dose estimates, while those allowing for 

broad variability along streamlines increased the frequency of "bottlenecks" reducing dose estimates. On this basis, 

simplified approaches with limited consideration of variability may suffice for intended uses of the performance 

assessment model, such as evaluation of site safety. 

4)� 40203 – Development of a Radiolytic Model for the Alteration of Spent Nuclear Fuel. Incorporation of 

non-oxidative matrix dissolution and hydrogen oxidation inhibition effect 

Lara Duro, Alba Valls, Olga Riba, Jordi Bruno, Amphos XXI Consulting S.L.(Spain); 

Aurora Martinez-Esparza, ENRESA (Spain) 

In the last years, there have been numerous efforts from national waste management agencies to develop models 

able to predict the dissolution behaviour of spent nuclear fuel under interim and/or long-term storage conditions. One of 

the most evolved models is the so called Matrix Alteration Model (MAM), which is based on the radiolytic oxidative 

dissolution of UO2 (Martínez-Esparza et al., 2004) and which has been applied to different experimental results with 

certain level of agreement. <strong>The</strong> calibration of the MAM model in front of new experimental data has resulted in the 

identification of some important drawbacks that may result in limited applicability of the model as a predictive tool. In 

this work we present the modifications made to the MAM in order to improve it and expand its range of applicability: - 

Incorporation of the non oxidative alteration of the matrix. Ignoring the incorporation of the non-oxidative alteration of 

the matrix has proved to be non-conservative over long-term experiments, resulting in an underestimation of the actual 

concentrations of uranium and radionuclides dissolving congruently with the matrix (Bruno et al., 2009). <strong>The</strong> 

incorporation of this mechanism in the MAM has been done by considering the different rates of irradiated and 

unirradiated UO2+x determined under reducing conditions and published in the open scientific literature. <strong>The</strong> modified 

MAM is able to reproduce experimental data gathered under a diverse range of experimental conditions. Incorporation of 

the catalytic effect of the surface on hydrogen activation. Hydrogen can be generated by radiolysis of water in the vicinity 

of the spent nuclear fuel as well as by anaerobic corrosion of metallic components forming on the container of the fuel 

under storage conditions. <strong>The</strong> ability of molecular hydrogen to decrease the rate of dissolution of the spent fuel matrix 

has been attributed to the presence of metallic surfaces in the fuel that can act as catalyst for the activation of molecular 

to atomic hydrogen (Carbol et al., 2009). <strong>The</strong> MAM model integrates the H2 inhibition effect on UO2 matrix oxidation 

by including reactions between H2 and the UO2 oxidising species, (i.e. consumption of oxidants species by H2) (Duro et 

al., 2009) but until now the activation of hydrogen on the surface of the solids present in the system had not been 

implemented. In this work we present the implementation of the heterogeneous activation of hydrogen in order to 

reproduce experimental conditions and couple the predictions of the long-term rates of fuel dissolution of the matrix. 

5)� 40172 – Evaluated and Estimated Solubility of Some Elements for Performance Assessment of Geological 

Disposal of High-level Radioactive Waste Using Updated Version of <strong>The</strong>rmodynamic Database 

Akira Kitamura, Reisuke Doi, JAEA (Japan); Yasushi Yoshida, Inspection Development Co., Ltd. (Japan) 

A thermodynamic database was established for performance assessment of geological disposal of high-level 

radioactive waste (HLW) and TRU waste (equated to long-lived intermediate level waste) based on the thermodynamic 

database (JNC-TDB) developed by the Japan Nuclear Cycle Development Institute in 1999. Twenty-five elements 

(actinides, fission products, activated nuclides and their progenies) which were important for the performance assessment 

of geological disposal were selected. <strong>The</strong> fundamental plan was in principle based on the guidelines established by the 

Nuclear Energy Agency (NEA) in the Organisation for Economic Co-operation and Development (OECD). Additional 

unique guidelines were established due to a requirement from the performance assessment to select tentative 

thermodynamic data obtained from chemical analogues and/or models for elements and species with insufficient 

thermodynamic data. <strong>The</strong>rmodynamic data for nickel, selenium, zirconium, technetium, thorium, uranium, neptunium, 

plutonium and americium, which were critically reviewed by the NEA, were taken to our thermodynamic database and 

modified. <strong>The</strong>rmodynamic data for cobalt and molybdenum, which were important for the performance assessment of 

TRU waste, were newly reviewed, selected and compiled. Some of thermodynamic data for other elements were updated 

or modified. All thermodynamic data were extrapolated zero ionic strength using the model called “specific ion 

interaction theory (SIT)”. We tried to assign uncertainty of the selected data as many as possible. Selected 

thermodynamic data were compiled as the JAEA’s thermodynamic database (JAEA-TDB), which were available for 

some geochemical calculation programs, e.g., PHREEQC. We evaluated and estimated solubility of the 25 elements in 

the simulated pore waters established in the second progress report (H12) for safety assessment of geological disposal of 

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Abstracts 

HLW in Japan using the JAEA-TDB and compared with the solubility values evaluated and estimated using the 

JNC-TDB. Furthermore, we tried to establish a technique to determine the solubility limiting solid for all the elements of 

interest. It was found that most of the evaluated and estimated solubility values were not changed drastically, but the 

solubility values and dominant aqueous species for some elements were changed using the JAEA-TDB, e.g., due to 

introducing the formation constant of polynuclear hydrolysis species of zirconium and replacing the formation constant 

of mixed carbonatohydoxo complexes of thorium. Detail of the comparison and discussion about the evaluated and 

estimated solubility values between the JAEA- and the JNC- TDBs will be presented. 

6) 40049 – Consideration on Soil Origin Carbon Transfer to Leafy Vegetables Using Stable Carbon Isotope 

Ratios 

Keiko Tagami, Shigeo Uchida, National Institute of Radiological Sciences (Japan) 

From the viewpoint of radiation dose assessment for humans from transuranic waste disposal sites, carbon-14 

(half-life: 5730 y) in organic forms is thought to be one of the most important radionuclides. Thus, understanding the 

C-14 fate in soil-to-crop systems is important since C-14 might transfer from waste disposal sites to crops through the 

soil environment. It is well known that carbon dioxide in the air is the major source for carbon assimilation by plants, 

thus, carbon in soil is less considered for plant uptake. Recent results by radiotracer experiments, however, indicated 

some contribution of soil origin carbon to plant carbon. In this study, we used stable carbon isotope ratios (C-13/C-12) in 

crop and soil samples. It is well known that C-13 and C-12 are fractionated in the photosynthesis process at a certain ratio, 

depending on plant types, i.e. C3, C4 and CAM plants. If all the carbon in crops originated only from the air, the stable 

carbon isotope ratio would take a constant value; however, if there were any contributions from soil carbon, the ratio 

would be changed. This approach has been used in our previous study for rice (Tagami and Uchida, 2009) and we tried to 

use the method for leafy vegetables. Leafy vegetables were cabbage, Chinese cabbage, lettuce, and leeks, and associated 

soil samples were also collected in Japan. Stable carbon isotope ratios and total C concentrations using an elemental 

analyzer connected to an isotope ratio mass spectrometer (<strong>The</strong>rmo Fisher Scientific, Flash EA and Delta V Advantage). 

Finally we found that the percentage of soil origin carbon to the total plant carbon was 2.9%, and the soil-to-plant 

transfer factors ranged 0.12-1.1 with an average of 0.5, which was within the previously reported values for radish 

(Sheppard et al., 1991). 

7) 40050 – Comparison of Soil-to-plant Transfer Factors for Rice and Wheat Grains 

Shigeo Uchida, Keiko Tagami, National Institute of Radiological Sciences (Japan) 

It is important to understand the behaviors of radionuclides in the agricultural systems since radionuclides enter 

humans through ingestion of foods. Critical staple foods are rice and wheat; the former is the dominant staple food crop 

in humid tropical and sub-tropical countries across the globe, and the latter is cultivated worldwide. Both are members of 

the Poaceae Family; however, they are different in cultivation methods. For example, rice is produced under flooded 

conditions but wheat is grown under upland field conditions. <strong>The</strong> difference in cultivation methods may affect plant 

uptake of radionuclides from soil. Thus we measured soil-to-grain transfer factors (TFs) of naturally existing elements as 

analogues of radionuclides for rice and wheat cultivated in Japan. For this work, we collected from 63 sites for rice and 7 

sites for wheat; to the results of wheat, we added 2 barley samples since their cultivation methods were the same. <strong>The</strong> 

elements we measured for TFs were Li, N, Na, Mg, Al, Si, P, K, Ca, Sc, Ti, V, C, Mn, Fe, Co, Ni, Cu, Zn, Ga, As, Se, Br, 

Rb, Sr, Y, Zr, Nb, Mo, Ag, Cd, Sn, Sb, I, Cs, Ba, La, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Hf, W, Tl, Pb, 

Ra, Th and U. Geometric means of TFs of these elements were compared between rice and wheat grains; the tendency 

was almost the same and the correlation factor was more than 0.98 by Student’s t-test (p

Abstracts 

hydrologic structure of the fault zone. A fourth borehole is planned to penetrate the main fault believed to lie in-between 

the holes. <strong>The</strong> main philosophy behind our approach for the hydrologic characterization of such a complex fractured 

system is to let the system take its own average and monitor a long term behavior instead of collecting a multitude of data 

at small length and time scales, or at a discrete fracture scale and to “up-scale,” which is extremely tenuous. 

2) 40189 – An attempt to evaluate horizontal crustal movement by geodetic and geological approach in the 


Tetsuya Tokiwa, Koichi Asamori, Tadafumi Niizato, Tsuyoshi Nohara, JAEA (Japan); 

Yuki Matsuura, Hitachi Zosen Co., Ltd. (Japan); Hideki Kosaka, Kankyo Chishitsu Co., Ltd. (Japan) 

In this study, we present the preliminary results for the estimation of a horizontal crustal movement by using 

geodetic and geological approach in the Horonobe area, northern Hokkaido, Japan. <strong>The</strong> estimations have been carried out 

by using a GPS data and a geological cross section obtained by applying balanced-section method. As results of this 

study, both of the shortening rates estimated by GPS data and balanced-section method indicate several millimeters per 

year. Namely, there is no contradiction between geodetic and geological data, and it is considered that Horonobe area is 

still situated similar tendency and magnitude of a crustal movement. It is seemingly considered that geodetic data is 

unhelpful for estimating the long-term crustal movement, because period of geodetic observations is a very short. 

However, the result of this study indicates that geodetic data provide valuable information for estimating the long-term 

crustal movement in the area, and it is considered that geodetic approach play an important role in improvement of the 

credibility of evaluation for prediction of long-term stability. 

3) 40054 – Relationship between hypocentral distribution and geological structure in the Horonobe area, 

northern Hokkaido, Japan 

Tetsuya Tokiwa, Koichi Asamori, Naoto Hiraga, Osamu Yamada, Hideharu Yokota, JAEA (Japan); 

Hirokazu Moriya, Tohoku University (Japan); 

Hikaru Hotta, Itaru Kitamura, Construction Project Consultants, Inc. (Japan) 

In this paper, we discuss the relationship between the accurate hypocentral distribution and three-dimensional (3-D) 

geological structure in and around the Horonobe area in northern Hokkaido, Japan. <strong>The</strong> multiplet-clustering analysis was 

applied to the 421 micro-earthquakes which occurred from 1 September, 2003 to 30 September, 2007. <strong>The</strong> 3-D 

geological structure model was mainly constructed from previous seismic reflection profiles and borehole data. As a 

result of this analysis, although with slight differences in depth between them, the hypocenters were found to be 

distributed in the NNW-SSE direction and become deeper from the west toward the east. <strong>The</strong> distributed pattern of the 

hypocenters is similar to that of the geological structure. <strong>The</strong>se results indicate that the hypocentral distribution may 

represent existence of active zone related to the geological structure, and provide effective information which can 

contribute to establishing methods for estimating the future evolution of the geological environment. 

4) 40062 – Technical know-how for modeling of geological environment (1) Overview and groundwater flow 

modeling 

Hiromitsu Saegusa, Shinji Takeuchi, Keisuke Maekawa, Hideaki Osawa, Takeshi Semba, JAEA (Japan) 

It is important for site characterization projects to manage the decision-making process with transparency and 

traceability and to transfer the technical know-how accumulated during the research and development to the 

implementing phase and to future generations. <strong>The</strong> modeling for a geological environment is to be used to synthesize 

investigation results. Evaluation of the impact of uncertainties in the model is important to identify and prioritize key 

issues for further investigations. <strong>The</strong>refore, a plan for site characterization should be made based on the results of the 

modeling. <strong>The</strong> aim of this study is to support for the planning of initial surface-based site characterization based on the 

technical know-how accumulated from the Mizunami Underground Research Laboratory Project and the Horonobe 

Underground Research Laboratory Project. <strong>The</strong>se projects are broad scientific studies of the deep geological environment 

that are a basis for research and development for the geological disposal of high-level radioactive wastes. In this study, 

the work-flow of the groundwater flow modeling, which is one of the geological environment models, and is to be used 

for setting the area for the geological environment modeling and for groundwater flow characterization, and the related 

decision-making process using literature data have been summarized. 

5) 40066 – Technical Know-how for Modeling ff Geological Environment (2) Geological Modeling 

Toshiyuki Matsuoka, Kenji Amano, Hideaki Osawa, Takeshi Semba, JAEA (Japan) 

It is important for site characterization projects to manage the decision-making process with transparency and 

traceability and to transfer the technical know-how developed and accumulated during the research and development to 

the implementing phase as well as to future generations. <strong>The</strong> modeling of a geological environment supports efforts to 

clarify the degree of understanding regarding that geological environment, including uncertainty. Evaluation of the 

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Abstracts 

impact of uncertainties in a geological environment model is important to identify and prioritize key issues for further 

investigations. <strong>The</strong>refore, a plan for site characterization should be made based on the results of the modeling. <strong>The</strong> aim of 

this study is to support the planning of initial surface-based site characterization based on the technical know-how 

accumulated from the Mizunami Underground Research Laboratory Project and the Horonobe Underground Research 

Laboratory Project. <strong>The</strong>se projects are broad scientific studies of the deep geological environment that are a basis for 

research and development for the geological disposal of high-level radioactive wastes. In this study, the work-flow 

followed in developing the geological model, one of the geological environment models, and the related technical 

know-how acquired from literature data have been summarized. 

6) 40039 – <strong>The</strong> long-term stability of geological environments in the various rock types in Japan from the 

perspective of uranium mineralization 

Eiji Sasao, JAEA (Japan) 

Long-term stability of the geological environment is one of the important keys for deep geological disposal of 

high-level radioactive waste in the Japanese Islands due to their location in a tectonically active island-arc. Uranium 

occurrences in Japan have been subjected to many geological processes inherent to the island-arc setting. Geological 

environments associated with uranium mineralization are considered favorable for HLW disposal, because uranium 

mineralization is considered a natural analogue of the radionuclides in HLW. Studies on the long-term stability of the 

uranium mineralization in Japan can be instructive as these could provide useful information on the long-term stability of 

the geological environment. Information on host rock and mode of occurrence of uranium mineralization was compiled 

from published data. <strong>The</strong> mineralization occurs in these types of deposits, i.e., sedimentary formations, association with 

metallic ore mineralization of magmatic origin and stratiform manganese mineralization, pegmatite, and alluvial placer 

deposit. <strong>The</strong> mineralization occurs in various geological settings in Japan. This fact suggests that geological 

environments suitable for geological isolation are widely distributed in the Japanese Islands, despite their location in a 

geologically active area. This study will support building confidence in HLW disposal in the Japanese Islands. 

SESSION D5: Measurement and Estimation 

1) 40045 – Improvement of Radioactivity Inventory Evaluation Procedure In Preparatory Tasks for 


Ken-ichi Tanaka, Hidenori Tanabe, Hideaki Ichige, JAPC (Japan) 

Preparatory tasks for decommissioning of nuclear power plant start with radiological characterization. Residual 

radioactivity inventory evaluation is a main part of the characterization. Reliable information on the inventory is 

important for specification for decommissioning plan. Japan Atomic Power Company (JAPC) has already started these 

tasks for Tsuruga Nuclear Power Plant Unit 1 (TS-1). We can optimize decommissioning plan using the information. To 

obtain the reliable information, we improved an evaluation procedure. <strong>The</strong> procedure is divided into two main steps. First 

step is neutron flux distribution calculation and second one is radioactivity distribution calculation. Radioactivity 

distribution is calculated using neutron flux distribution. In this work, we improved the evaluation procedure to obtain the 

reliable information on the inventory Because of the limitation of computer resource, two-dimension (2D) approximation 

model was applied to radioactivity distribution around Reactor Pressure Vessel (RPV). We can calculate reliable 2D 

neutron flux distribution by having better understanding of neutron transport phenomena. Neutron flux was measured at 

30 locations in TS-1 Primary Containment Vessel (PCV) using activation foils. And in order to understand the neutron 

transport phenomenon inside the PCV, we also calculated neutron flux distribution with the three-dimensional (3D) 

discrete ordinates method calculation (Sn) code. By consideration about the result of the measurement and 3D calculation, 

we could understand the characteristics of the neutron flux distribution inside the PCV. To simulate the neutron flux 

distribution well with 2D Sn code, neutron flux behaviors inside the PCV had been investigated with referencing the 

measurement values and with observing calculated 3D neutron flux distribution. 2D calculation model had been modified 

repeatedly until reliable calculation result was provided. After several model modifications, the reliable 2D calculation 

was accomplished and important neutron transport phenomena that are necessary to simulate the neutron flux distribution 

well was understood. Network-parallel-computing technique was applied to radioactivity distribution calculation. Using 

this technique, we could calculate radioactivity at all space mesh points that were used with 2D Sn code and we obtained 

the radioactivity distribution. By using this distribution, we can estimate a quantity of radioactivity around RPV more 

accurately and optimize dismantling designs. 

2) 40202 – Verification of Source Term Analysis System for Decommissioning Wastes from CANDU Reactor 

Dong-Keun Cho, Gwang-Min SUN, Jongwon Choi, KAERI (Korea Rep.); 

Donghyeun Hwang, Hak-Soo Kim, Tae-Won Hwang, KHNP (Korea Rep.) 

<strong>The</strong>re are now twenty commercial nuclear power reactors operating as of May 2010 in South Korea. As nuclear 

capacity becomes higher and installations age, the Korean government and industry have launched R&D to estimate 

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Abstracts 

appropriate decommissioning costs of power reactors. In this paper, MCNP/ORIGEN2 code system which is being 

developed as a source term evaluation tool was verified by comparing the estimated nuclide inventory from 

MCNP/ORIGEN2 simulation with the measured nuclide inventory from chemical assay in an irradiated pressure tube 

discharged from Wolsong Unit 1 in 1994. Equilibrium core model of Wolsoung unit 1 was used as a neutron source to 

activate in-core and ex-core structural components. As a result, the estimated values from the analysis system agreed with 

measured data within 20% difference. <strong>The</strong>refore, it can be concluded that MCNP/ORIGEN system could be a reliable 

tool to estimate source terms of decommissioning wastes from CANDU reactor, although this system assumes constant 

flux irradiation and snapshot equilibrium core model as a reference core. 

3) 40294 – Evaluation of <strong>The</strong> Activated Radioactivity of Turbine Equipments in BWR 

Masato Watanabe, Motonori Nakagami, Chubu Electric Power Co., Inc. (Japan) 

<strong>The</strong> equipments in the main steam system under the plant operation were irradiated by the neutron emitted from 

N-17 which is produced by (n,p) reaction of the coolant in Boiling Water Reactor (BWR) plants. However, only few 

measures, regarding the radioactive concentration of N-17 in the main steam system of BWR, were indicated up to now. 

In this study, the radioactive concentration of N-17 in the main steam system of the advanced boiling water reactor 

(ABWR) plant, Hamaoka unit-5, is evaluated by the neutron flux measurements and calculations. 

It was found that the concentration of N-17 at the reactor pressure vessel outlet nozzle of the main steam pipe is 

approximately 3 Bq/cm3. It turns out that the concentration has small dependency on nuclear thermal power and that it is 

relatively common among the BWR plants. 

<strong>The</strong>n, the activated activities of turbine equipments and pipes in the main steam system are evaluated by using the 

measured radioactive concentration of N-17. <strong>The</strong> activities of the equipments in the main steam system are relatively 

very low level. 

4) 40014 – Optimization of Quantitative Waste Volume Determination Technique for Hanford Waste Tank 

Closure 

Yi Su, David L. Monts, Ping-Rey Jang, Zhiling Long, Walter P. Okhuysen, Olin P. Norton, Lawrence L. Gresham, 

Jeffrey S. Lindner, Mississippi State University (USA) 

<strong>The</strong> Hanford Site is currently in the process of an extensive effort to empty and close its radioactive single-shell and 

double-shell waste storage tanks. Before this can be accomplished, it is necessary to know how much residual material is 

left in a given waste tank and the uncertainty with which that volume is known. 

<strong>The</strong> Institute for Clean Energy Technology (ICET) at Mississippi State University is currently developing a 

quantitative in-tank imaging system based on Fourier Transform Profilometry, FTP. FTP is a non-contact, 3-D shape 

measurement technique. By projecting a fringe pattern onto a target surface and observing its deformation due to surface 

irregularities from a different view angle, FTP is capable of determining the height (depth) distribution (and hence 

volume distribution) of the target surface, thus reproducing the profile of the target accurately under a wide variety of 

conditions. Hence FTP has the potential to be utilized for quantitative determination of residual wastes within Hanford 

waste tanks. In this paper, efforts to characterize the accuracy and precision of quantitative volume determination using 

FTP and the use of these results to optimize the FTP system for deployment within Hanford waste tanks are described. 

5) 40120 – Implementation of Decommissioning Materials Conditional Clearance Process to the OMEGA 

Calculation Code 

Matej Zachar, Vladimir Necas, Slovak University of Technology (Slovakia); 

Vladimir Daniska, DECONTA a.s. (Slovakia) 

<strong>The</strong> activities performed during nuclear installation decommissioning process inevitably lead to the production of 

large amount of radioactive material to be managed. Significant part of materials has such low radioactivity level that 

allows them to be released to the environment without any restriction for further use. On the other hand, for materials 

with radioactivity slightly above the defined unconditional clearance level, there is a possibility to release them 

conditionally for a specific purpose in accordance with developed scenario assuring that radiation exposure limits for 

population not to be exceeded. <strong>The</strong> procedure of managing such decommissioning materials, mentioned above, could 

lead to recycling and reuse of more solid materials and to save the radioactive waste repository volume. In the paper an 

implementation of the process of conditional release to the OMEGA Code is analyzed in details; the Code is used for 

calculation of decommissioning parameters. <strong>The</strong> analytical approach in the material parameters assessment, firstly, 

assumes a definition of radiological limit conditions, based on the evaluation of possible scenarios for conditionally 

released materials, and their application to appropriate sorter type in existing material and radioactivity flow system. 

Other calculation procedures with relevant technological or economical parameters, mathematically describing e.g. final 

radiation monitoring or transport outside the locality, are applied to the OMEGA Code in the next step. Together with 

limits, new procedures creating independent material stream allow evaluation of conditional material release process 

during decommissioning. Model calculations evaluating various scenarios with different input parameters and 

considering conditional release of materials to the environment are performed to verify the implemented methodology. 

Output parameters and results of the model assessment are presented, discussed and concluded in the final part of the 

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paper. 

Abstracts 

6) 40183 – Quantitative determination of the initial components in the activated pressure tubes of the Wolsong 

1st CANDU reactor 

Gwang-Min Sun, Dong-Keun Cho, KAERI (Korea Rep.) 

When the end of the original lifespan is coming for the Wolsong 1st CANDU reactor in Korea, decommissioning 

has been one of the largest issues requiring large expense, which is facing the nuclear industry and the government. 

Korean government intends to provide public funds, which have been accumulated using a portion of the electrical 

charge, to dismantle reactors when the time comes. <strong>The</strong> radioactive sources in the construction stuffs and components of 

reactor such as pressure tube, steam generator and so on must be evaluated for the estimation of appropriate expense for 

the decommissioning. <strong>The</strong> overall objective of this study is to confirm the methods for the elemental or isotopic analysis 

of the pressure tube samples from nuclear reactor in order to make possible the reliable application of the methods to the 

case of the Wolsong 1st CANDU reactor. To achieve this objective we need to categorize the methods into two groups 

taking into account the applicability of the methods to the target we have interests in. <strong>The</strong> radiochemical analytical 

methods such as liquid scintillation counting, X-ray spectroscopy and normal gamma-ray spectroscopy and so on to 

measure the trace isotopic beings such as 60Co, 94Nb, 54Mn, 55Fe, 59Ni, 63Ni and so on. Elemental analysis methods 

such as instrumental neutron activation analysis (INAA) and prompt gamma activation analysis (PGAA) to measure the 

initial compositions of the samples. In present work, we will deal with INAA and PGAA to determine an elemental 

composition of the pressure tube. <strong>The</strong> prepared samples are irradiated in the NAA #1, NAA #2 and SNU-KAERI PGAA 

facility of the HANARO research reactor in Korea Atomic Energy Research Institute. Because the samples are from 

pressure tube of reactor, the most abundant element must be iron and other elements can be determined relative to the 

iron content. From this idea, we determined the initial elemental composition of the samples through two-step procedure. 

For the absolute determination of iron content in the sample, the count rate must be calibrated according to the iron 

elemental content as shown in figure 8 where the slope of the line is a specific count rate or analytical sensitivity given in 

cps/?g. <strong>The</strong> relative mass of other elements in the samples were determined by using a k0-standardization method which 

has been usually used in the INAA and also in the PGAA. 

SESSION R3: ER Techniques 

1) 40286 – Sequential Extraction and Determination of Depleted Uranium in the Presence of Natural Uranium in 

Environmental Soil samples by ICP-MS 

Mohamed Amr, Alaa E. Negmeldin, Khalid Al-Saad, A. T. Al-Kinani, Qatar University (Qatar); 

A. I. Helal, Atomic Energy Authority (Qatar) 

Determination of depleted uranium (DU) in the presence of natural uranium (NU) by inductively coupled plasma 

mass spectrometry (ICP-MS) was applied on environmental soil samples collected from Qatar. <strong>The</strong> soils were artificially 

spiked by soaking them in a mixture of DU and NU. <strong>The</strong> detection limit of 235U and 238U isotopes were 141 ppt and 

1.28 ppt, respectively. It was observed from sequential experiment that uranium was brought into solution mainly 

appeared at steps for dilute acid-soluble, carbonate-bound and organic matter-bound species. Little redistribution was 

observed at steps for exchangeable and Fe-Mn oxide-bound species. 

2) 40096 – Determination of Environmental Uranium Concentration by Utilizing Gamma-ray Emission from the 

Progeny Radionuclides 

Tadao Tanaka, Taro Shimada, Takenori Sukegawa, JAEA (Japan); Takeshi Ito, Japan ATOX Co., Ltd. (Japan) 

Nuclear facility sites such as enrichment plant and fabrication plant are allowed to be released from nuclear safety 

regulations after the plants are decommissioned. <strong>The</strong> sites are necessary to confirm to be decontaminated, prior to be 

released. 

Gamma-ray emission from the progeny radionuclides of uranium such as Th-234, Pa234m, Ra-226 has been utilized 

for the determination of uranium concentration in soils. Gamma-ray emission radionuclides occurring in spacious areas 

of land was often measured by the in-situ method with portable germanium semiconductor detector (portable Ge 

detector), to confirm that there is no significant gamma-ray emission radionuclides distributed in the vast land areas. In 

the present study, we proposed a determination method for U-238 concentration of background level in environment and 

for probate of vast site areas, in which the gamma-ray radioactivities from Th-234, Pa-234m, Ra-226 are measured with 

the portable Ge detector. 

Validity of the estimation method of U-238 concentration from the progeny radionuclides was examined by the 

comparison between the U-238 concentration estimated by the in-situ method with portable Ge detector and that directly 

measured by ICP-MS. <strong>The</strong> U-238 concentration by the in-situ method was estimated from peak counting rate at 63 keV 

of the gamma-ray emission corresponding to Th-234 and from that at 186 keV corresponding to Ra226. <strong>The</strong> estimated 

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Abstracts 

U-238 concentration was in the order of 0.01 Bq/g in radioactive concentration, and was in comparable level with the 

concentrations decided by the ICP-MS. 

<strong>The</strong> determination method of U-238 concentration from the progeny radionuclides was applied to the site 

contaminated by low-level uranium resourced from an uranium handling facility. <strong>The</strong> U-238 concentration could be 

determined from the peak counting rates corresponding to Th-234 and Ra-226. <strong>The</strong> proposed method utilizing 

gamma-ray radioactivities from the progeny radionuclides may be available for the U-238 concentration determination in 

vast land areas. 

3) 40170 – Natural Radioactivity Levels And Gamma-Ray Dose Rate In Soil And Transfer Of Radionuclides 

From Soil To Vegetation Of Some Northern Area Of Pakistan Using Gamma-Ray Spectrometry 

Muhammad Ismail, Hasan M.Khan, Peshawar University (Pakistan); Khalid Khan, Pinstech Islamabad (Pakistan) 

<strong>The</strong> analysis of gamma emitters natural radionuclides i.e. 226Ra, 232Th and 40K has been carried out in soil, 

vegetation, vegetable and water samples collected from some Northern area of Pakistan, using gamma-ray spectrometry. 

<strong>The</strong> gamma-ray spectrometry was carried out using high purity Germanium (HPGe) detector coupled with a computer 

based high resolution multi channel analyzer. <strong>The</strong> activities concentrations in soil ranged from 24.7 to 78.5 Bq kg-1, 21.7 

to 75.3 Bq kg-1 and 298.5 to 570.8 Bq kg-1 for 226Ra, 232Th and 40K with the mean value of 42.1, 43.3, 9.5 and 418.3 

Bq kg-1, respectively. <strong>The</strong> radium equivalent activity, internal and external hazard indices have mean values of 136.11 

Bq kg-1, 0.48 and 0.37, respectively. <strong>The</strong> mean values of outdoor and indoor absorbed dose rate in air were found 64.61 

nGy.h-1 and 77.54 nGy.h-1, respectively. In the present analysis, 40K was the major radionuclide present in soil, 

vegetation, fruit and vegetable samples. <strong>The</strong> activity concentration of 40K in vegetation sample varied from 646.6 Bq 

kg-1 to 869.6 Bq kg-1 on dry weight basis. However, the activity concentration of 40K in fruit and vegetable samples 

varied from 34.0 Bq kg-1 to 123.3 Bq kg-1 on fresh weight basis. <strong>The</strong> transfer factors of these radionuclides from soil to 

vegetation and vegetable were also studied. <strong>The</strong> analysis of water samples showed activities concentrations for all 

radionuclides below detection limit. <strong>The</strong> activities concentrations of radionuclides found in all samples during the current 

investigation were nominal. <strong>The</strong>refore they are not associated with any potential source of health hazard to the general 

public. 

4) 40034 – Effect of Fertilizer and Soil Amendments on Extraction Yields of Radioiodine and Radiocesium in Soil 

Hirofumi Tsukada, Akira Takeda, Shunichi Hisamatsu, Institute for Environmental Sciences (Japan) 

Mobile fraction of radionuclides in a soil, which is water-extractable and exchangeable, is important information for 

remediation in the terrestrial environment. Radioiodine and radiocesium derived from atmospheric nuclear weapons tests 

and released from nuclear facilities are major radionuclides for the assessment of radiation exposure to the public. In the 

present study, the effects of fertilizer and soil amendments on the extraction yields of radioiodine and radiocesium in a 

soil were investigated. A surface soil (Andosol) was collected from grassland in Aomori, Japan. <strong>The</strong> soil was mixed with 

the following amendments: Chemical fertilizer (NH4, PO4, K): 0.24 mg N, 0.16 mg P2O5, or 0.16 mg K2O per g soil, 

Compost (rice straw mixed with cattle feces): 50 mg per g soil, and Clay mineral (illite, kaolinite, montmorillonite, 

sericite, vermiculite, zeolite): 20 mg per g soil. 

<strong>The</strong>y were mixed and then 10 kBq of 125I or 137Cs was added to one g of the soil sample. <strong>The</strong>y were stored in an 

artificial climate chamber, and a wetting-and-drying treatment was repeated by adding one ml of deionized water every 2 

weeks. <strong>The</strong> extraction yields of 125I with water, and 137Cs with water or 1 M ammonium acetate solution were 

determined at 10, 30 and 120 d after addition. <strong>The</strong> extraction yield of 125I in the soil without amendment (the control 

sample) decreased to 4% of the added amount at 120 d after addition. <strong>The</strong> extraction yield of 125I in the soil mixed with 

the compost at 10 and 30 d after addition was higher than that in each control soil, and the difference was accompanied 

by an increase of the dissolved organic carbon content in the water extract of the compost application soil. <strong>The</strong> extraction 

yield of 125I in the soil mixed with the montmorillonite was higher than that in the control soil through the aging period. 

<strong>The</strong> extraction yield of 137Cs in the control soil with water and ammonium acetate decreased to 0.12% and 32% at 120 d 

after addition, respectively. <strong>The</strong> extraction of 137Cs with water was enhanced by the application of the nitrogen fertilizer 

and the compost, whereas that was depressed by the application of the clay minerals. <strong>The</strong> extraction of 137Cs with 

ammonium acetate was also depressed by the clay minerals, and this may be caused by the large cesium interception sites 

in the clay minerals. It was clear that the mobile fractions of radioiodine and radiocesium in the soil were changed by 

several treatments. This work was conducted under contract with the Aomori Prefectural Government, Japan. 

5) 40246 – Impact of Mobile-Immobile Water Domains on the Retention of Technetium (Tc-99) in the Vadose 

Zone 

Danielle Jansik, Dawn Wellman, Elsa Cordova, PNNL (USA) 

<strong>The</strong> transport of technetium (Tc-99), like many other radionuclides, is of interest due to the potential for human 

exposure and impact on ecosystems. Technetium has been released to the environment through nuclear power production 

and nuclear fuel processing; as a result, further spreading of Tc-99 is a concern at DOE sites across the US. Specifically, 

technetium is a contaminant of concern at Hanford, Savannah River, Idaho, and Oak Ridge National Laboratory. <strong>The</strong> 

current body of work conducted on Tc-99 has provided a wealth of information regarding the redox relationships, 

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Abstracts 

sorption, solubility, and stability of the mineral phases (Artinger et al., 2003; Beals and Hayes, 1995; Cui and Eriksen, 

1996b; Gu and Schulz, 1991; Jaisi et al., 2009; Keith-Roach et al., 2003; Kumar et al., 2007), however little work has 

been conducted on the physical transport of the highly soluble pertechnetate oxyanion, in the subsurface. Current 

conceptual models do not explain the persistence and presence of the anion in deep vadose zone environments such as the 

Hanford site. In an oxic reducing environment with low organic content the residence time of technetium is the soil 

would be expected to be near low, due to its low sorption. Surprisingly, nearly 50 years following the release of 

contamination into the site, much of the element has persisted in the subsurface, in its most mobile form. Using an 

Unsaturated Flow Apparatus (UFA) we have conducted a series of experiments to examine the impact of 

mobile-immobile domains on the transport of Tc-99. By varying sand/silt ratios and saturations we examined how 

changes in pore geometry and moisture content impact the transport of Tc-99 within our experimental system. Results 

demonstrating the impact of sediment texture pore morphology, and soil moisture content on physical impediments to 

Tc-99 transport will be presented. 

6) 40207 – Nano Technologies for Environmental Remediation 

Harch Gill, Lehigh Nanotech (USA) 

This paper presents two innovative nanotechnology products, NanoFeTM and Nano-OxTM, that have been 

successfully used for the in-situ treatment of soil and groundwater. NanoFeTM consists primarily of submicron (

Abstracts 

SESSION L7: Storage and Disposal Facility 

1) 40095 – Design, Licensing and Commissioning Activities for the Waste Storage Facility to Manage the Waste 

Streams from Retubing of Wolsong-1 CANDU 600 MW Nuclear Generating Station in South Korea 

Nikos Pontikakis, John Ballantine, Sharon Hugh, Joulien Moussalam, AECL (Canada) 

Until recently, the operating license for the Waste Storage Facility of a CANDU 600 MW Nuclear Generating 

Station was limited to meeting the operational requirements for the initial 25 to 30 years of station life. However, as of 

April 2009, the Wolsong-1 station in South Korea was shut down for a Plant Life Extension project, which included 

retubing of the reactor core. <strong>The</strong> retube work generated large volumes of low and intermediate radioactive retubing waste 

streams with specific characteristics that typically cannot be handled and stored within the existing capacity of the Waste 

Storage Facility at Wolsong site. 

Atomic Energy of Canada Limited has identified and assessed the requirements for handling and storing the 

radwaste streams generated from retubing a CANDU 600 MW plant. At Wolsong-1 CANDU 600 MW Nuclear 

Generating Station, these requirements are met by expanding the existing Waste Storage Facility, design and supply of 

retubing waste handling equipment and waste containers. 

<strong>The</strong> extension to the existing Solid Radioactive Waste Management Facility includes construction of retube waste 

canisters. <strong>The</strong>se concrete storage structures are cooled by natural air convection and store two types of waste containers 

for intermediate level waste such as pressure tubes, calandria tubes, calandria tube inserts and end fittings. 

Low level waste that is generated by retubing is comprised of: o Reactor components that were exposed to low 

neutron flux and contain activated corrosion product deposits (for example feeder pipe sections, positioning assembly 

hardware, closure plugs, etc) o Retubing engineered consumables from tooling that cannot be reused for future retubing 

projects, most of which is generated from the following systems: § Retube Active Ventilation System, and § Volume 

Reduction System. 

<strong>The</strong> low level waste streams are packaged in Type A boxes for safe transportation of the waste. <strong>The</strong> low level waste 

in Type A boxes are stored in the existing Low Level Waste Facility at Wolsong site. 

This paper describes, at a high level, the following aspects of the design, licensing and commissioning activities of 

the Waste Storage Facility to support retubing of the Wolsong-1 CANDU 600 MW Nuclear Generating Station in South 

Korea: 

Ø Retubing waste management design requirements; Ø Design of retube waste canisters, waste handling equipment, 

waste containers for intermediate level waste and Type A boxes for low level waste Ø Licensing process; and Ø 

Commissioning activities. 

2) 40284 – Microbial Occurrence in Bentonite-based Buffer Materials of a Final Disposal Site for Low Level 

Radioactive Waste in Taiwan 

Fong-In Chou, Chia-Chin Li, Tzung-Yuang Chen, National Tsing Hua University (Taiwan); 

Hsiao-Wei Wen, National Chung Hsing University (Taiwan) 

This research addresses the potential of microbial implications in bentonite for use as a buffer and backfill material 

in final disposal site for low-level radioactive waste (LLRW) in Taiwan, where has a special island-type climate. Microbe 

activities naturally present in this site were analyzed, and buffer materials (BM) consisted of 100%, 70% or 50% 

bentonite were prepared for laboratory studies. A total of 39 microbial strains were isolated, and the predominant strains 

included four bacterial, one yeast and four fungal strains. Growth inhibition was not detected in any tested strain cultured 

in a radiation field with a dose rate of 0.2 Gy/h. Most of the isolated strains grew under a dose rate of 1.4 Gy/h. <strong>The</strong> D10 

values of the tested strains ranged from 0.16 to 2.05 kGy. <strong>The</strong> mycelia of tested fungal strains could spread over 5 cm 

during six months of inoculation in BM. <strong>The</strong> spreading activity of the tested bacteria was less than that of the fungi. 

Moreover, biofilms were observed on the surfaces of the BM. Since a large and diverse population of microbes is present 

in Taiwan, microbes may contribute to the mobilization of radionuclides in the disposal site. 

3) 40153 – Assessing the gas transport mechanisms in the Swiss L/ILW concept using numerical modeling 

Irina Gaus, Paul Marschall, Joerg Rueedi, Nagra (Switzerland); 

Rainer Senger, John Ewing, Intera Inc. Swiss Branch (Switzerland) 

In low/intermediate-level waste (L/ILW) repositories, anaerobic corrosion of metals and degradation of organic 

materials produce hydrogen, methane, and carbon dioxide. Gas accumulation and gas transport in a L/ILW repository is 

an important component in the safety assessment of proposed deep repositories in low-permeability formations. <strong>The</strong> 

dominant gas transport mechanisms are dependent on the gas overpressures as with increasing overpressure the gas 

transport capacity of the system increases. <strong>The</strong> dominant gas transport mechanisms occurring with increasing gas 

pressure within the anticipated pressure ranges are: diffusion of gas dissolved in pore water (1), two phase flow in the 

hostrock and the excavation damaged zone (EDZ) whereby no deformation of the pore space occurs (2), gas migration 

within parts of the repository (if repository materials are appropriately chosen) (3) and pathway dilation (4). Under no 

circumstances the gas is expected to induce permanent fractures in the hostrock. This paper focuses on the gas migration 

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in parts of the repository whereby materials are chosen aimed at increasing the gas transport capacity of the backfilled 

underground structures without compromising the radionuclide retention capacity of the engineered barrier system (EBS). 

<strong>The</strong>se materials with enhanced gas permeability and low water permeability can supplement the gas flow that is expected 

to occur through the EDZ and the host rock. <strong>The</strong> impact of the use of adapted backfill and sealing materials on the gas 

pressure build-up and the major gas paths were assessed using numerical two-phase flow models on the repository scale. 

Furthermore, both the gas and water fluxes as a function of time and gas generation rate can be evaluated by varying the 

physical properties of the materials and hence their transport capacity. Results showed that by introducing seals with 

higher gas permeability, the modeled gas flow is largely limited to the access tunnels and the excavation disturbed zone 

for the case of a very low permeability host rock. <strong>The</strong> bulk of the gas flows through the repository seal and the adjacent 

EDZ into the tunnel system. In addition to the demonstration of the gas flow in the seal and access tunnel system by 

numerical models, laboratory results confirm the high gas transport capacity of the sand/bentonite mixtures. In a next step 

a multi year demonstration scale experiment (GAST) at the Grimsel Test Site is envisioned. 

4) 40283 – <strong>The</strong> progress and results of Demonstration Test of Cavern-Type Disposal Facility 

Yoshihiro Akiyama, Kenji Terada, Nobuaki Oda, Tsutomu Yada, Takahiro Nakajima, RWMC (Japan) 

<strong>The</strong> cavern-type disposal facilities for low-level waste (LLW) with relatively high radioactivity levels mainly 

generated from power reactor decommissioning and for part of transuranic (TRU) waste mainly from spent fuel 

reprocessing are designed to be constructed in a cavern 50 to 100 meters below ground, and to employ an engineered 

barrier system (EBS) of a combination of bentonite and cement materials in Japan. In order to advance the feasibility 

study for these disposal, a government-commissioned research project named Demonstration Test of Cavern-Type 

Disposal Facility started in fiscal 2005, and since fiscal 2007 a full-scale mock-up test facility has been constructed under 

actual subsurface environment. <strong>The</strong> main objective of the test is to establish construction methodology and procedures 

which ensure the required quality of the EBS on-site. By fiscal 2009 some parts of the facility have been constructed, and 

the test has demonstrated both practicability of the construction and achievement of the quality. <strong>The</strong>y are respectively 

taken as low-permeability of less than 5x10-13 m/s and low-diffusivity of less than 1x10-12 m2/s at the time of 

completion of construction. This paper covers the project outline and the test results obtained by the construction of some 

parts of a bentonite and cement materials. 

SESSION H9: Repository Engineering and Demonstration 

1) 40304 Keynote – Repository engineering and demonstration: special challenges for TRU 


<strong>The</strong> diverse range of long-lived radioactive wastes without significant heat output specified for deep geological 

disposal (here termed TRU) pose challenges that are potentially more serious than those from vitrified high-level waste 

and spent fuel. Despite this, the latter tend to be the focus of R&D in national programmes. Such challenges are 

particularly severe for the case for countries that are not considering evaporite host rocks or have a volunteering approach 

to siting and those with inventories of TRU resulting from reprocessing of spent fuel. While there is little doubt that safe 

disposal of TRU is feasible, it is tricky to develop a convincing safety case for a site during early stages of 

characterisation as, compared to HLW/SF, less credit can be taken for robust, long-term performance of current designs 

of the engineered barrier systems. In order to improve this situation and increase flexibility with respect to host rock 

properties, two different options are available – improving the conditioning of particular waste streams or improving the 

overall repository safety concept. Although the former has been a focus for work in some countries (particularly Japan), 

much less effort has been invested in the latter and hence this will be illustrated by some examples. In the paper these 

options will be compared in terms of their pros and cons with respect to practicality of implementation, environmental 

impact and cost. Additionally, the ease with which the resulting safety case can be supported by demonstrations of key 

arguments will be discussed, which may indicate the likely degree of acceptance by stakeholders. 

2) 40119 – Half-Scale Test: An important step to demonstrate the feasibility of the Belgian Supercontainer 

concept for disposal of HLW 

Lou Areias, SCK CEN and VUB (Belgium); 

Bart Craeye, Ghent University/Artesis Hogeschool Antwerpen (Belgium); 

Geert De Schutter, Ghent University (Belgium); 

Hughes Van Humbeeck, William Wacquier, ONDRAF/NIRAS (Belgium); 

Alain Van Cotthem, Loic Villers Technum-Tractebel (Belgium) 

This paper presents results of a half scale test performed by ESV EURIDICE, an Economic Interest Grouping 

between the Belgian Nuclear Research Centre (SCK•CEN) and the Belgian Agency for Radioactive Waste and Enriched 

Fissile Materials (ONDRAF/NIRAS). <strong>The</strong> primary objective of the test was to assess the feasibility of constructing the 

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Supercontainer and to provide experimental data to validate modelling calculations obtained using the finite element 

program HEAT/MLS. <strong>The</strong> test focused on the early age behaviour of the concrete matrix materials and the practical 

aspects of construction. Generally, the results obtained from the half scale test confirm that it is feasible to construct the 

Supercontainer with currently available techniques. <strong>The</strong> results also validate scoping calculations obtained earlier with 

the finite element model. <strong>The</strong>se findings contribute an important step to demonstrate the feasibility to construct the 

Supercontainer and to validate the Belgian Supercontainer concept proposed by ONDRAF/NIRAS for disposal of high 

level waste (HLW) in Belgium. 

Keywords: Belgian supercontainer concept, high-level radioactive waste, half-scale test, concrete buffer, overpack, 

construction feasibility 

3) 40175 – Full-Scale Test on Overpack Closure Techniques for HLW Repository Operation - Welding Methods 

and UT Systems For Long-Term Structural Integrity of the Weld Joint - 

Ario Nakamura, Hidekazu Asano, RWMC (Japan); 

Takashi Furukawa, JAPAN POWER ENGINEERING AND INSPECTION CORPORATION (Japan); 

Kyosuke Fujisawa, KOBE STEEL, LTD. (Japan); Susumu Kawakami, IHI Corporation (Japan); 

Takashi Ito, Mitsubishi Heavy Industries, Ltd. (Japan). 

Overpack, a high-level radioactive waste package for Japan’s geological disposal program, is required for preventing sealed 

vitrified waste from contact with groundwater for at least 1,000 years. <strong>The</strong> weld joint between the body and lid must also meet this 

requirement. Certain welding methods were examined for applicability through full-scale welding tests using various welding depths 

up to 190 mm and two different lid structures. Results show that generation of welding flaws must be considered unavoidable. 

<strong>The</strong>refore, ultrasonic testing (UT) must be conducted on the assumption that weld flaws will be present. Such UT systems must be 

designed for natural defects. Several types of UT must be evaluated for detection and size estimation capability at depths ranging from 

the surface to the bottom of the weld joint. Certain UT methods were examined for their ability to detect natural defects that were 

created on the surface of and inside a 190-mm thick carbon steel specimen. Probability of detection (POD) of each UT method was 

calculated by comparing the results of UT and destructive examination. In consideration of the preferred range of scanning depth for 

each UT method, a concept that combines UT methods was proposed as a practical UT system for the overpack weld joint. 

KEYWORDS: radioactive waste management, geological disposal, waste package, overpack, final closure weld, structural 

integrity, natural defects, UT, POD 

4) 40242 – Design Options for HLW Repository Operation Technology, (I) Demonstration and Evaluation of 

Remote Handling Technologies 

Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan); Hideki Kawamura, Obayashi Corporation (Japan) 

<strong>The</strong> long-term safety of a geological repository for high-level radioactive waste(HLW) in Japan is achieved using a 

multi-barrier system, with emphasis on an engineered barrier system (EBS) consisting of bentonite-based buffer material 

and a steel overpack containing vitrified wastes. Providing a wide variety of technical options for repository operation, 

which can allow EBS construction under various repository conditions, is valuable for the volunteering approach to siting 

a HLW repository in Japan. In order to confirm technological applicability of candidate techniques and to evaluate 

compatibility in an operation system, full-scale tests of various remote-handling and-emplacement techniques for the 

EBS were conducted. <strong>The</strong> test results are summarized in the technical menu which is hierarchical structured database. 

Since the long-term performance of EBS is significantly influenced by remote handling and emplacement techniques for 

EBS, consistency of the operation technology must be evaluated from the view point of system standing. In this program, 

evaluation methodology was studied and several indexes, representing requirements for the repository operation and 

management, were identified. Indices include current technical availability, long-term safety of the resultant EBS, 

operational safety, engineering reliability, retrievability, etc., with expert assessment complementing a more quantitative 

evaluation. 

5) 40251 – Design Options for HLW Repository Operation Technology, (II) Bentonite Block Forming and 

Vertical Emplacement/JGC 

Hajime Takao, Tatsuhiro Takegahara, JGC Corporation (Japan); 

Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan) 

RWMC and JGC have been running an all-round R&D program for the period of 2000-2010 to develop the concept 

of Vertical Emplacement for disposal of vitrified waste. <strong>The</strong> conceptual design of its basic equipment was worked out in 

2000, followed by forming the large-scale bentonite block in 2001-2004. Study has also been conducted on a mechanism 

to convey and position the large-scale block using a vacuum suction device. Subsequent to these developments, various 

technologies necessary for designing the Vertical Emplacement equipment have been reviewed, which would enhance 

engineering feasibility and reliability. 

Full-scale demonstration program under a joint research program with JAEA (Japan Atomic Energy Agency) 

started in 2008 with the twin objectives i) supporting of public relations and ii) technical verification. <strong>The</strong> large-scale 

bentonite block and part of the full-scale Vertical Emplacement equipment are now on view at the Full-scale 

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demonstration facility in Horonobe, Hokkaido, Japan. 

Abstracts 

6) 40268 – Design Options for HLW Repository Operation Technology, (III) Transportation and Horizontal 

Emplacement of Pre-Fabricated EBS Module (PEM) 

Susumu Kawakami, IHI Corporation (Japan); Hitoshi Nakashima, Hidekazu Asano, RWMC (Japan) 

As one of the repository operation technologies for high-level radioactive waste?(HLW), the pre-fabricated 

engineered barrier system (EBS) module (PEM) was carried out the examination of handling and emplacement technique 

for EBS. <strong>The</strong> PEM technology was examined to confirm technological applicability. <strong>The</strong> PEM is concept of the 

integration of EBS as the module in the surface facilities, and transporting the module underground facilities. This 

concept is the one of the candidate concepts of horizontal emplacement techniques for EBS in Japan. <strong>The</strong>refore, PEM is 

the same level large size and heavy weight as EBS, and it is necessary to examine the applicability of handling and 

emplacement techniques. Full-scale level tests were performed to confirm the applicability of these techniques with the 

air bearing/air jack devices. In the tests, we prepared the testing devices of full-scale level size/weight and confirmed the 

applicability of these technologies as an elemental technology on the condition of considering the environment of an 

underground tunnel. <strong>The</strong> air bearing test that produced the surface-roughness of the tunnel environment was carried out 

the evaluation concerning the transportation performance of the air bearing. And, the air jack test was carried out the 

holding and emplacement of PEM. <strong>The</strong> repository operation technology with the air bearing/jack device was confirmed 

to execute the examination, and to apply to handling and emplacement technique for PEM. 

7) 40236 – Design Options for HLW Repository Operation Technology, (IV) Shotclay Technique for Seamless 

Construction of EBS 

Ichizo Kobayashi, Soh Fujisawa, Makoto Nakajima, Masaru Toida, Kajima Corporation (Japan); 


In Japan, the construction method of the buffer material has been investigated focusing on the block emplacement 

and in-situ compaction methods. Under the current concept of geological disposal of radioactive waste, since it has been 

important that the barrier satisfies the dry density requirement as a mass, no enough attention has been paid to the 

distribution of density and the gaps between blocks which were caused by the conventional methods. This is based on the 

assumption that bentonite swells from the permeation of groundwater and that the density eventually becomes 

homogeneous. However, it is not clear whether the high density bentonite does in fact swell until the density is 

homogeneous. <strong>The</strong> effects of density distribution and the gaps in long-term high impermeability of a 

bentonite-engineered barrier may be not small. <strong>The</strong>refore, the option of the construction method for a 

bentonite-engineered barrier that does not have such uncertainty concerning the implementation of a disposal concept 

should be prepared. In order to remove such uncertainty, the high-density spray method for bentonite, termed the 

shotclay method, was developed as a method for constructing a uniform bentonite-engineered barrier without gaps even 

in a narrow space. <strong>The</strong> shotclay method is used to construct compacted soil at high density by spraying material 

preconditioned with water. Using this method, the dry density of 1.6 Mg /m3, which was considered impossible with the 

spraying method, is achieved. In this study, the applicability of the shotclay method to HLW bentonite-engineered 

barriers was confirmed experimentally. In the tests, an actual scale vertical-type HLW bentonite engineered barrier was 

constructed. This was a bentonite-engineered barrier with a diameter of 2.22 m and a height of 3.13 m. <strong>The</strong> material used 

was bentonite with 30% silica sand, and water content was adjusted by mixing chilled bentonite with powdered ice 

before thawing. Work progress was 11.2 m3 and the weight was 21.7 Mg. <strong>The</strong> dry density of the entire buffer was 1.62 

Mg/m3, and construction time was approximately 8 hours per unit. After the formworks were removed, the core and 

block of the actual scale HLW bentonite-engineered barrier were sampled to confirm homogeneity. As a result, 

homogeneity was confirmed, and no gaps were observed between the formwork and the buffer material and between the 

simulated waste and the buffer material. <strong>The</strong> applicability to HLW of the shotclay method has been confirmed through 

this examination. 

8) 40254 – Design Options for HLW Repository Operation Technology, (V) Preliminary Study and Small Scale 

Experiments on the Method of Removal of Buffer Material with Salt Solution 

Satohito Toguri, Jiho Jang, Takashi Ishii, Mitsunobu Okihara, Kengo Iwasa, SHIMIZU CORPORATION (Japan); 


During the construction of geological disposal facilities for high-level radioactive waste, it may be decided to free 

and retrieve the emplaced overpack for some reasons. Thus we have been paying attention to a method of slurrying 

bentonite buffer around the overpack with fluid (salt solution) for freeing it. A few laboratory tests were performed to 

research the feasibility and verify the applicability of the method. <strong>The</strong> test piece of the buffer material consists of 70wt% 

bentonite and 30wt% sand, and its dry density is 1.6 g/mL, and the volume of the cylindrical test piece is 100mL. First, 

the time was measured for test pieces to be dissolved in various strength NaCl solutions. In case of 47% saturated test 

pieces, they could be easily dissolved in NaCl 3 or 4wt% solution, and the dissolved material was deposited soon in the 

slurry. Second, the results from the flushing experiments applied to a small specimen of buffer material suggest that 

flushing 4wt% NaCl solution is effective for speedy stripping of buffer material. Third, the times were measured for 

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various percentage saturated test pieces to be dissolved in 4wt% NaCl solution. <strong>The</strong> time for 47% saturated one was 

shorter than 1 Hr, but the time for 67% was longer than 5 Hrs. In case of 88%, it had not been dissolved yet in 7Hrs, but 

after drying the test piece it could be easily dissolved in 5 minutes. <strong>The</strong> results from the laboratory tests indicate that the 

unsaturated buffer material can be easily dissolved in NaCl 4wt% solution, and the wet buffer material can be dissolved 

by immersing in NaCl 4wt% solution after process of dry-hot-air blowing. Finally, in order to confirm the effect of 

removal processes, simulated experiments were executed using small scale (1/14 scale) specimens. <strong>The</strong> 47% saturated 

buffer material specimen was easily removed by the process of immersing in NaCl solution, flushing NaCl solution, 

removing slurry by vacuum device, and re-using NaCl solution after deposition. In the nearly saturated case, it was 

removed in cyclic process of dry-hot-air blowing and immersing in NaCl 4wt% solution. 

SESSION H10: Site Characterization and Modeling of Geological Environment (2) 

1) 40135 – Dry-run of Site Investigation Planning using the Manual for Preliminary Investigation in Japan 

Shigeki Akamura, Tadashi Miwa, NUMO (Japan); Tatsuya Tanaka, Obayashi Corporation (Japan); 

Hiroshi Shiratsuchi, Tokyo Electric Power Service Co.,Ltd. (Japan); 

Atsushi Horio, DIA CONSULTANTS CO., Ltd. (Japan) 

A stepwise site selection process has been adopted for geological disposal of HLW in Japan. Literature surveys, 

followed by preliminary investigations (PI) and, finally, detailed investigations in underground facilities will be carried 

out in the successive selection stages. In the PI stage, surface-based investigations such as borehole surveys and 

geophysical prospecting will be implemented. 

NUMO recognizes that sustained improvement of internal expertise is very important to ensure that the PI will be 

implemented rigorously and efficiently. <strong>The</strong>refore existing knowledge and experience in the planning and management 

of site investigations were compiled in the form of two manuals: the Preliminary Investigation Planning Manual (PIPM) 

and the Preliminary Investigation Management Manual (PIMM). <strong>The</strong> manuals were based on the experience in overseas 

site investigation programs, and has further been refined by taking experience from site investigations in Japan (e.g. from 

generic URLs) into account. NUMO has applied them in its own R&D programs, such as the dry-run studies for various 

geological environments (e.g. coastal and island). 

This paper outlines the process and the results of a dry-run study which applied the revised PIPM to the Yokosuka 

area where the demonstration and validation project for PI technologies has been conducted. <strong>The</strong> planning of the PI was 

performed according to the eight steps in the PIPM. <strong>The</strong> GET (Geosphere Evaluation Team) consisting of staffs from site 

characterization, performance assessment and repository design groups in NUMO, was established for the PI planning. 

This task force team has responsibility not only for the planning but also for making key decisions regarding the 

performance of the program. <strong>The</strong> SDMT (Site Descriptive Modeling Team) was formed for site modeling activities. 

Initial geoscientific conceptual models were established based on the interpretation of existing site information around 

the area. Various conceptual models were constructed due to the limited amount of existing information. Site descriptive 

models were derived taking such variations in the conceptual models into account. <strong>The</strong> main targets for the PI planning 

were specified considering the uncertainties in the models, requests from the repository concept as well as legal 

requirement. <strong>The</strong> integrated PI program set out the sequence of investigations and characterization activities to achieve 

the main targets. Lessons learned in the PI planning and the applicability of the revised PIPM were addressed as the 

conclusion of the dry-run. 

2) 40070 – Evaluation of the long-term evolution of the groundwater system in the Mizunami area, Japan 

Takashi Mizuno, Teruki Iwatsuki, JAEA (Japan); Antoni E. Milodowski, British Geological Survey (UK) 

This study aimed to develop a methodology for assessing the evolution of the long-term groundwater system, using 

fracture-filling calcite. Fracture-filling calcite mineralization, closely associated with groundwater flow paths, in deep (to 

ca. 1000 m) granitic rocks in and around Mizunami area, Japan, was studied by optical and cathode-luminescence 

microscopy, SEM, laser ablation microprobe-ICP MS, stable isotope geochemistry, and fluid inclusion analysis. As the 

results, four generations (I to IV) of calcite precipitation can be differentiated based on their morphological and isotopic 

characteristics. Calcite I - calcite with indistinct morphology including the wall rock fragments (d13CPDB -31.4 ‰ ~ 

-7.5 ‰, d18OPDB -32.7 ‰ ~ -8.0 ‰); Calcite II -calcite with euhedral rhombohedral and hexagonal crystal forms 

(d13CPDB 22.5 ‰ ~ -1.8 ‰, d18OPDB -15.5 ‰ ~ -6.5 ‰); Calcite III - calcite forming c-axis elongated rhombohedral 

overgrowths seeded on top of Calcite II (d13CPDB -18.0 ‰ ~ 5.8 ‰, d18OPDB -11.1 ‰ ~ -2.3 ‰); Calcite IV - calcite 

forming small rhombohedral crystals nucleated on the surface of Calcite III (d13CPDB -12.5 ‰ ~ -1.7 ‰, d18OPDB 

-12.7 ‰~7.8 ‰). Carbon and oxygen isotopic ratios suggests that the Calcite I is of hydrothermal origin. On the other 

hand, Calcite II, IV and III were precipitated from freshwater and marine water, respectively. From the geohistorical 

point of view, depositional setting in Mizunami area was changed during Tertiary and Quaternary. Mizunami Group 

overlying above Toki granite is changed from lacustrine strata to marine strata. In addition, upper Seto Group is a 

lacustrine stratum. <strong>The</strong> change of depositional setting corresponds to groundwater system assumed by mineralogical 

study. As there is no other evidence linked to penetration of high-salinity water into deep environment in this area, it is 

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suggested that both of deposition of marine strata (upper part of Mizunami group) and forming Calcite III precipitation 

were possibly caused by same transgression event. After Calcite III precipitation, calcite IV was precipitated from fresh 

water during flushing of the marine water. As summary, integrated morphological, mineralogical, microchemical and 

isotopic analysis of multilayered calcite fracture mineralization associated with groundwater flow paths provides valuable 

information to evaluate long-term evolution of groundwater system. 

3) 40077 – Study on the Estimation Error Caused by Using One-dimensional Model for the Evaluation of Dipole 

Tracer Test 

Yuji Ijiri, Yumi Naemura, Taisei Corporation (Japan); 

Kenji Amano, Keisuke Maekawa, Kunio Ota, Takanori Kunimaru, Atsushi Sawada, JAEA (Japan) 

In-situ tracer tests are of value in obtaining parameters for repository performance assessment. A one-dimensional 

model is simple and has thus been commonly employed to identify the radionuclide transport parameters by fitting the 

model to breakthrough curves obtained from the tracer tests. It can, however, be considered that the one-dimensional 

model could increase uncertainty in the identified parameters. In particular, such errors are not negligible when the 

parameters are evaluated for the test conducted in a dipole (twodimensional) flow field between injection and pumping 

wells. In this study, the effects of various experimental conditions including pumping rate, dipole ratio, heterogeneity of 

fracture transmissivity and background groundwater flow on the identified parameters are investigated using computer 

simulations for the case of tracer tests in a fracture plane. Longitudinal dispersivity tends to be overestimated by using 

the one-dimensional model and to become larger when the pumping rate becomes smaller, the dipole ratio larger, the 

heterogeneity of the fracture stronger and orthogonal oriented background groundwater flow greater. Such information 

will be of much help for planning tracer tests in an appropriate manner at underground research laboratories both at 

Mizunami in central Japan and Horonobe in northern Japan. Definition of appropriate experimental conditions will 

contribute to decreasing the uncertainty in the results of the tracer tests. 

4) 40056 – Development of Comprehensive Techniques for Coastal Site Characterisation (1) Strategic Overview 

Kunio Ota, Kenji Amano, Tadafumi Niizato, JAEA (Japan); 

W Russell Alexander, Bedrock Geosciences (Switzerland); 

Yoshiaki Yamanaka, Suncoh Consultants (Japan) 

<strong>The</strong> assurance of the long-term stability of the geological environment is sine qua non for deep geological disposal. 

Any assessment of repository safety will thus require development of a set of analyses and arguments to demonstrate the 

persistence of the key safety functions of the geological environment up to several hundred thousand years into the future, 

taking into account the likely future evolution of the repository host rock. 

Concern presently focusses on a sea-level rise caused by anthropogenic global warming but, within a period of 

several tens of thousands of years, a return to glacial-period conditions is to be expected. Based on previous glaciations, 

global sea-level could drop by up to 150 m and, in the case of Japan, such a decrease in sea-level would result in an 

extremely dramatic change in the location of coastal lines with a subsequent significant change to hydraulic and 

hydrochemical conditions at coastal sites. It is thus of great importance in the Japanese disposal programme to establish 

comprehensive techniques for characterising the overall evolution of coastal sites over geological time with focus very 

much on the persistence of the key safety functions throughout episodes of uplift/subsidence and climatic and sea-level 

changes. 

To this end, based on practical experience from the ongoing underground research laboratory projects of JAEA, a 

transparent and traceable roadmap for planning and implementing a sequence of field investigations at any coastal site 

has been formulated. Known as a “Geosynthesis Data Flow Diagram (GDFD)”, this system illustrates linkages between a 

range of parameters, including investigations of key aspects to be addressed, interpretation of data acquired, synthesis of 

the results of different studies and analyses and final clarification of the key properties and processes of the geological 

environment. In particular, the GDFD defines a geosynthesis methodology for describing temporal and spatial changes of 

various characteristics and processes, with particular focus on the site palaeohydrogeology. 

Such a geosynthesis methodology has been introduced in an ongoing collaborative programme for characterising 

the coastal geological environment around Horonobe in northern Hokkaido, Japan. A basic strategy for stepwise 

investigations has been proposed, which incorporates the geosynthesis methodology in an effective manner in each step 

from initial survey/review of existing information, through aerial, terrestrial and marine exploration, to the final borehole 

programme. This technique has now been tested and optimised based on technical findings and experience that have been 

accumulated with the progress of the investigations. 

5) 40052 – Development of comprehensive techniques for coastal site characterisation: (3) Conceptualisation of 

long-term geosphere evolution 

Tadafumi Niizato, Kenji Amano, Kunio Ota, Takanori Kunimaru, JAEA (Japan); 

Lanyon Bill, Nagra (Switzerland); 


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A critical issue for building confidence in the long-term safety of geological disposal is to demonstrate the stability 

of the geosphere, taking into account its likely future evolution. This stability is broadly defined as the persistence of 

<strong>The</strong>rmal-Hydrological-Mechanical-Chemical conditions considered favourable for the long-term safety of a geological 

repository. <strong>The</strong> geosphere is slowly but constantly evolving, and then the stability, in this case, does not imply that 

steady-state conditions exist. What is important is that the evolution of the geosphere can be understood. In general, an 

understanding of the evolution is gained by studying the palaeohydrogeological evolution of a site, defining temporal and 

spatial changes of various characteristics, events, and processes over geological time. <strong>The</strong> site palaeohydrogeology refers 

to natural events and processes that have occurred in the past and contributed to the present state of the geosphere, which 

include sub-surface processes (e.g. crustal movement, diagenesis, etc.) and earth-surface processes (e.g. climatic and 

sea-level changes, geomorphological processes, etc.). An understanding of the palaeohydrogeological evolution of the 

site provides the firm foundation to describe the likely future evolution of the site. An ongoing collaborative programme 

aims to establish comprehensive techniques for characterising the overall evolution of coastal sites through studying the 

palaeohydrogeological evolution in the coastal system around the Horonobe area, Hokkaido, northern Japan. In this study, 

the current status for the conceptualisation of the long-term geosphere evolution in the coastal area, is based on data from 

the JAEA’s underground research laboratory project. Information on surface and sub-surface processes has been 

integrated into a chronological conceptual model which indicates space-time sequences of the events and processes in the 

area over geological time. Spatial scale for the conceptualisation is ca. 100 km in the East-West direction through the 

locations of the underground research laboratory and of the borehole investigations on the coast in the Horonobe area. 

Temporal conceptualisations over the last few million years are focused on the spatial and temporal changes of the 

geosphere caused by sub-surface processes and over the last several hundred thousand years focused on the changes 

caused by earth-surface processes. <strong>The</strong> methodology for conceptualisation of the geosphere evolution will be applied to 

other analogous coastal areas on Japan’s western seaboard to produce comprehensive techniques to support 

understanding the geosphere evolution of potential coastal sites for deep geological repositories. 

6) 40048 – Development of Comprehensive Techniques for Coastal Site Characterisation (2) Integrated 

Palaeohydrogeological Approach for Development of Site Evolution Models 

Kenji Amano, Tadafumi Niizato, Hideharu Yokota, Kunio Ota, JAEA (Japan); 

Bill Lanyon, Nagra (Switzerland); 


Radioactive waste repository designs consist of multiple safety barriers which include the waste form, the canister, 

the engineered barriers and the geosphere. In many waste programmes, it is considered that the three most important 

safety features provided by the geosphere are mechanical stability, favourable geochemical conditions and low 

groundwater flux. To guarantee that a repository site will provide such conditions for timescales of relevance to the 

safety assessment, any repository site characterisation has to not only define whether these features will function 

appropriately today, but also to assess if they will remain adequate up to several thousand to hundreds of thousand years 

into the future, depending on the repository type. 

In general, this is done by studying the palaeoKHNPgeological evolution of a site, defining temporal and spatial 

changes of various characteristics and processes. <strong>The</strong>se may include hydrogeology, geology, groundwater flow 

characteristics, groundwater chemistry and site tectonics, including uplift and erosion processes. <strong>The</strong>se key aspects are 

studied to build up a conceptual model for the overall site evolution over geological time, up to the present and this is 

used to define the likely future evolution of the site and to assess if the main safety features will continue to function 

adequately. 

<strong>The</strong> collaborative programme described here is focussed on the palaeohydrogeology of the coastal area around 

Horonobe in northern Hokkaido, Japan. Data from JAEA’s ongoing underground research laboratory project is being 

synthesised in a Site Descriptive Model (SDM) with new information from the collaborating research institutes to 

develop a Site Evolution Model (SEM), with the focus very much on changes in the Sea of Japan seaboard over the last 

few million years. This new conceptual model will then be used to assess the palaeohydrological evolution of other 

analogous sites on Japan’s western seaboard, with the final aim of producing a set of comprehensive techniques to 

understand the palaeohydrogeological evolution of the deep geosphere of all coastal sites on the Sea of Japan. 

7) 40041 – Development of Methodology of Groundwater Flow and Solute Transport Analysis in the Horonobe 

Area, Hokkaido, Japan 

Keisuke Maekawa, Hitoshi Makino, Hiroshi Kurikami, Tadafumi Niizato, Manabu Inagaki, 

Makoto Kawamura, JAEA (Japan) 

It is important for establishment of safety assessment techniques of geological disposal to understand groundwater 

flow and solute transport accurately. <strong>The</strong>refore, we are positioning to confirm an applicability of the techniques in 

realistic environment as a crucial issue in R&D. We have attempted and planed some relevant studies as below: -A 

methodology to integrate activities from site investigations to evaluation of solute transport was examined. We have 

carried out groundwater flow analysis on a regional scale using geological and hydrological information from 

surface-based investigations at the Horonobe area, and also solute transport analysis based on the information of the 

trajectory analysis. -We have carried out a preliminary simulation of groundwater flow and salinity concentration 

distribution using information on climatic and sea-level changes, and evolution of geological structures considering the 

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impacts of natural events and processes. Consequently, we could outline the impacts of natural events and processes on 

geological environment including hydrogeology, hydrochemistry and their evolutions. -We have been planning to 

develop and apply a methodology of groundwater flow and solute transport analysis to the shallow part, the Horonobe 

coastal area and around the URL. <strong>The</strong>se techniques would become a basis for future site specific safety assessment in 

Japan. 

SESSION D6: Waste Treatment and Non-Reactor 

1) 40105 – Estimation of Radioactivity of Graphite Blocks in Tokai Power Station using Statistical Method 

Masaaki Nakano, Fuji Electric Holdings Co., Ltd. (Japan); 

Hisashi Mikami, Fuji Electric Systems Co., Ltd. (Japan); 

Hideaki Ichige, Shinich Tsukada, JAPC (Japan) 

Tokai Power Station (graphite moderated, gas-cooled reactor, GCR) stopped its commercial operation in March 

1998 and is decommissioning now. Since graphite blocks in Tokai reactor core are major low level wastes (LLWs), the 

realistic and reasonable method to estimate radioactivity of graphite blocks is required for final disposal and its licensing 

procedure. In general, LLWs, which were installed in or around a reactor core, have large radioactivity, theoretical 

calculations can be applied to the estimation of the radioactivity. This paper describes the concept of the method using 

statistical approach to determine the radioactivity of the graphite blocks in the reactor core. This method directly 

considers the variations of input calculation conditions, for example, compositions of impurity elements, irradiation 

neutron flux and irradiation period. In this paper, the variations of the compositions of impurity elements were 

statistically considered with the mean value and the standard deviation that were determined with elemental analyses. 

Many activation calculations were performed with the compositions that were determined with pseudorandom numbers, 

the mean value and the standard deviation. <strong>The</strong> calculated radioactivities distribute also statistically and a mean value and 

a standard deviation of radioactivity can be determined. <strong>The</strong> distribution of calculated radioactivities shows consistency 

to radiochemical analyses of graphite blocks from the reactor core and this shows that the method is applicable to the 

estimation of the graphite block radioactivity. Furthermore, this method can be considered to reduce over-excess 

estimation margin and can obtain reasonable radioactivity rather than using maximum or conservative values of all input 

conditions. This method is now being developed and approved as one of basic procedure for determining the 

radioactivity of wastes by Standards Committee of the Atomic Energy Society of Japan. 

2) 40115 – <strong>The</strong> treatment of hexavalent chromium in waste liquid from Fugen Decommissioning 

Yuki Yahiro, Seiji Yamamoto, Koji Negishi, Hitoshi Sakai, Tadashi Fukushima, 

Norimasa Yoshida, Toshiba Corporation (Japan); Nobuo Ishizuka, Yuji Sato, Wataru Fujiwara, JAEA (Japan) 

<strong>The</strong> prototype Advanced <strong>The</strong>rmal Reactor (ATR) Fugen Nuclear Power Station of JAEA terminated its operation in 

2003 and is under decommissioning. In cooling water (90m3) for iron-water shield concrete surrounding the core region, 

hexavalent chromium (K2CrO4) was dissolved as corrosion inhibitor. <strong>The</strong> hexavalent chromium has to be reduced to 

trivalent chromium, for the safety. Authors had developed a new treatment technique for the waste liquid containing 

hexavalent chromium with small amount of secondary waste. We manufactured and installed test equipment, and we 

treated the waste liquid in Fugen. We added formic acid as pH adjuster in the water and hydrogen peroxide to reduce 

hexavalent chromium to trivalent chromium. <strong>The</strong> residual hydrogen peroxide was decomposed to oxygen and water with 

iron catalysis and UV lamp. <strong>The</strong>n cationic species, such as trivalent chromium and potassium, in waste liquid were 

removed by cation exchange resin. <strong>The</strong> hydrogen peroxide and iron catalysis were added again to decompose the residual 

formic acid to carbon dioxide gas and water. Next, the residual hydrogen peroxide was decomposed by catalase. Finally, 

all the ionic species in the waste liquid were removed by mixed resin. It was confirmed that the concentration of 

hexavalent chromium could be reduced to below 0.5ppm and total chromium concentration to below 2.0ppm. This value 

is below the water pollution prevention law. It took 5 days for the whole treatment of 5m3 waste liquid containing the 

hexavalent chromium 340ppm. 

3) 40201 – Characterization of Radioactive Waste from Side Structural Components of a CANDU Reactor for 

Decommissioning Applications in Korea 

Rizwan Ahmed, Gyunyoung Heo, Kyung Hee University (Korea Rep.); 

Dong-Keun Cho, Jongwon Choi, KAERI (Korea Rep.) 

Reactor core components and structural materials of nuclear power plants to be decommissioned have been 

irradiated by neutrons of various intensities and spectrum. This long term irradiation results in the production of large 

number of radioactive isotopes that serve as a source of radioactivity for thousands of years for future. Decommissioning 

of a nuclear reactor is a costly program comprising of dismantling, demolishing of structures and waste classification for 

disposal applications. <strong>The</strong> estimate of radio-nuclides and radiation levels forms the essential part of the whole 

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decommissioning program. It can help establishing guidelines for the waste classification, dismantling and demolishing 

activities. ORIGEN2 code has long been in use for computing radionuclide concentrations in reactor cores and near core 

materials for various burn-up-decay cycles, using one-group collapsed cross sections. Since ORIGEN2 assumes a 

constant flux and nuclide capture cross-sections in all regions of the core, uncertainty in its results could increase as 

region of interest goes away from the core. This uncertainty can be removed by using a Monte Carlo Code, like MCNP, 

for the correct calculations of flux and capture cross-sections inside the reactor core and in far core regions. MCNP has 

greater capability to model the reactor problems in much realistic way that is to incorporate geometrical, compositional 

and spectrum information. In this paper the classification of radioactive waste from the side structural components of a 

CANDU reactor is presented. MCNP model of full core was established because of asymmetric structure of the reactor. 

Side structural components of total length 240 cm and radius 16.122 cm were modeled as twelve (12) homogenized cells 

of 20 cm length each along the axial direction. <strong>The</strong> neutron flux and one-group collapsed cross-sections were calculated 

by MCNP simulation for each cell, and then those results were applied to ORIGEN2 simulation to estimate nuclide 

inventory in the wastes. After retrieving the radiation level of side structural components of in- and ex-core, the 

radioactive wastes were classified according to the international standards of waste classification. <strong>The</strong> wastes from first 

and second cell of the side structural components were found to exhibit characteristics of class C and Class B wastes 

respectively. However, the rest of the waste was found to have activity levels as that of Class A radio-active waste. <strong>The</strong> 

waste is therefore suitable for land disposal in accordance with the international standards of waste classification and 

disposal. 

4) 40068 – Uranium refining and conversion plant decommissioning project 

Naoki Zaima, Yasuyuki Morimoto, Noritake Sugitsue, Kazumi Kado, JAEA (Japan) 

<strong>The</strong> uranium refining and conversion plant at Ningyo-toge (URCP) was constructed in 1981 for the purpose of 

demonstration on refining and conversion processes from yellow cakes to UF6 via UF4, and then as modified to develop 

the conversion of reprocessed uranium production of natural UF6 and purification of reprocessed UF6. Through 20 years, 

385 tons of natural uranium of UF6 and 336 tons of reprocessed uranium of UF6 had been conducted. <strong>The</strong>re are two 

refining processes in the URCP facilities. One is the wet-type process for the natural uranium and the other is the 

dry-type process for the reprocessed uranium. It was found the large amount of uranium residuals such as wet slurry and 

dried powder inside the vessels and pipings. Careful consideration had always been required against the diffusion of 

contamination. <strong>The</strong> basic policies concerning plant decommissioning are the optimization of the labor costs and the 

minimization of the radioactive wastes. <strong>The</strong> procedures are followings; i)measuring doserate by high sensitivity 

surveymeters and identificating nuclide by gamma ray spectrometry, ii)estimating uranium mass inventory, iii) planning 

workers distributions including of radiation control staffs, iv)deciding dismantling methods and decontaminating 

schematically if required, v)measuring and classifying doserate and contamination level, vi) managing for radioactive 

waste container, vii)control for personal exposures. Through two years and half, almost all equipment had been 

dismantled except building decontamination. Several hundreds tons of dismantled wastes had accumulated in 200 litter 

drums approximately. In addition, the secondary wastes had also been generated. Several thousands day of working time 

had spent totally. <strong>The</strong> radiation monitoring of working places had been performed during dismantling, the results were 

generally less than 20?Sv/h under the doserate limitation. However, followed by the trace of the reprocessed uranium, 

U-232 progenies nuclides such as Th-228 and Tl-208 were observed. <strong>The</strong> expected exposures are apprehensive by 

accumulation of the high energy gamma emission nuclides. For example, the fluidization media storage tank in which 

Th-228 progenies originated from U-232 accumulated. <strong>The</strong>re arises the case of remote controlled tools for decreasing 

personal exposure if required. As for the waste disposal, the determination of uranium content must be necessary. We are 

now developing for measuring systems with better accuracy. <strong>The</strong> further tasks imposed us summarized the followings; 

i)dismantling method for high doserate area, ii)reduction of radioactive wastes volume, iii)decontamination for the 

buildings, iv)waste disposal. 

5) 40161 – Radiochemical Characterization of Concrete Samples from the Environment of Accelerator Facilities 

Dorothea Schumann, Paul Scherrer Institute (Switzerland) 

<strong>The</strong> Paul Scherrer Institute (PSI) in Switzerland operates the most powerful proton accelerator facility in Europe 

and also world-wide, with a beam current of 2 mA and a final proton energy of 590 MeV. It consists of two injector 

cyclotrons with beam energies of 72 MeV each and a ring cyclotron accelerating the protons to the final energy. Besides 

the targets (mainly carbon wheels for the muon production and solid lead as the target for the neutron production) also 

the closer surroundings of the facility are highly activated. One of the components there is concrete which is used in huge 

amounts as construction material for walls and floor and acts as an effective and cheap shielding. Activation of these 

components could be neglected in the past because of the relatively low dose rates, but this situation changed with the 

upgrade of the accelerator during the last years. <strong>The</strong>refore, detailed declaration of the radionuclide content is also 

requested for concrete from the Swiss authorities (NAGRA). In the present contribution, we report on the radiochemical 

analysis of two different kinds of shielding concrete. <strong>The</strong> first samples stem from the floor of a former graphite target 

station, which was dismounted in 1989. Here we expect activation by secondary neutrons produced by the primary proton 

beam of 590 MeV from the ring cyclotron. <strong>The</strong> second series was taken from the environment of a copper target hit by 72 

MeV protons from one of the injector cyclotrons. <strong>The</strong>se samples were taken in 1998. <strong>The</strong> main activation is also in this 

case caused by the produced secondary neutrons. Besides conventional gamma-spectrometry, identifying mainly 

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lanthanides and 60Co as activation products, also long-lived radionuclides like 10Be, 26Al, 36Cl and 129I were found to 

be present. <strong>The</strong> results of both series are compared. Conclusions for a final disposal can be drawn from the obtained data. 

SESSION M2: Public Involvement 

1) 40288 – Activities of the OECD/NEA in the Field of Stakeholder Confidence for Radwaste Management and 



<strong>The</strong> OECD/NEA seeks to assist its member countries in developing strategies for the management of all types of 



facilities. <strong>The</strong> programme of work in the area of stakeholder involvement is supervised by the Radioactive Waste 

Management Committee (RWMC) and carried out by its Forum on Stakeholder Confidence (FSC). <strong>The</strong> latter is made of 

senior representatives from regulatory authorities, decommissioning organisations, policy making bodies, and 

research-anddevelopment institutions from the NEA countries. First and foremost, the FSC is a “learning organisation”. 

Through the FSC, members seek to improve themselves as responsive actors in the governance of radioactive waste 

management and decommissioning. Delegates attend in order to benefit from in-depth pragmatic exchanges with both 

peers and stakeholders beyond the membership. <strong>The</strong>y then consolidate their learning: the FSC takes as a responsibility to 

mature its lessons in discussion and cooperation with those concerned, and then to validate its conclusions with the help 

of academic researchers. <strong>The</strong> third step in the cycle is to make the learning available to others. <strong>The</strong> FSC has held 

workshops in six countries and local communities therein according to its own well tested approach. In each place, the 

host defines the principal themes for discussion in radioactive waste management. FSC members learn from national 

presenters about the history and context of each case study and hear a broad range of stakeholder voices describe their 

position, actions and concerns. Sitting together in small groups for roundtable discussions, FSC members can ask 

questions of hosting stakeholders, understand better their point of view, and share experience from their own institutions 

and countries. Each table then gives feedback to the entire audience, and the main observations are published alongside 

the texts of stakeholder presentations. <strong>The</strong> FSC then elaborates further on the lessons to be learnt. Over the 10 years of its 

existence, the FSC has established itself as a key international player and adviser in the field of stakeholder confidence. 

Its numerous studies and its advice are available on-line in the form, respectively, of academic reports and policy 

brochures and flyers. Explored topics include: stepwise decision-making, principles for stakeholders involvement, the 

partnering approach for siting and developing fuel cycle facilities, lessons learned in decommissioning, providing 

value-added in view of building a durable relationship between a facility and its host community, etc. Current study areas 

include: the interests and roles of regional authorities, how to increase the knowledge base of journalists, and providing 

added value (beyond economic benefits) to communities hosting waste management facilities. <strong>The</strong> activities and lessons 

to be learnt will be reviewed for the ICEM-2010 audience. 

2) 40219 – A Comparative Study of Stakeholder Participation in the Cleanup of Radioactive Wastes in the US, 

Japan and UK 

Mito Akiyoshi, Senshu University (Japan); William Lawless, Fjorentina Angjellari-Dajci, Paine College (USA); 

Christian Poppeliers, Augusta State University (USA); John Whitton, National Nuclear Laboratory (UK) 

We review case studies of stakeholder participation in the environmental cleanup of radioactive wastes in the USA, 

Japan and UK (e.g., [21,26,27,66,78]). Citizen participation programs in these three countries are at different stages: 

mature in the US, starting in Japan, and becoming operational in the UK. <strong>The</strong> US issue at the US Department of Energy’s 

(DOE) Savannah River Site (SRS) in South Carolina (SC) had been focused on citizens encouraging Federal (DOE; US 

Environmental Protection Agency, or EPA; and the US Nuclear Regulatory Commission, or NRC) and State (SC's 

Department of Health and Environmental Compliance, or DHEC) agencies to pursue "Plug-in-RODs" at SRS to simplify 

the regulations to accelerate closing seepage basins at SRS. In Japan, the Reprocessing of spent fuel and deep geological 

disposal of vitrified high-level waste have been among Japan's priorities. A reprocessing plant in Rokkasho, Aomori 

Prefecture is expected to commence operations in October 2010. <strong>The</strong> search of a site for a deep geological disposal 

facility has been ongoing since 2002. But the direct engagement of stakeholders has not occurred in Japan. Indirectly, 

stakeholders attempt to exert influence on decision-making with social movements, local elections, and litigation. In the 

UK, the issue is gaining effective citizen participation with the UK's Nuclear Decommissioning Authority (NDA). We 

hope that the case studies from these countries may improve citizen participation. 

3)� 40076 – Territorial Integration of the Geological Repository in France 

Gerald Ouzounian, Sebastien Farin, Roberto Miguez, Jean-Louis Tison, Andra (France) 

In France, a framework has been drawn up by the National Assembly and implemented by the government, in order 

to get the best relationship between Andra, among others, and the stakeholders and the inhabitants of the towns and 

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countries where disposal facilities or projects are or could be established. <strong>The</strong> main threads of the two Acts passed* in 

2006, being relevant to the relationship with inhabitants are the information exchange and the local economic 

development. Dealing with the information exchange and diffusion: - <strong>The</strong> Local Information Committee (CLI), for each 

nuclear facility, has been reinforced and a specific, Local Information and Oversight Committee (for the Underground 

Laboratory in Meuse-Haute-Marne) has been renewed. <strong>The</strong> CLI was in charge of a general assignment to inform and 

consult on nuclear safety, radioprotection and environmental topics. Now, since 2006, the nuclear facility’s CLI and the 

CLIS are able to order study reports, measures and analyses to experts freely selected. - Creation of the High Committee 

for Transparency and Information on Nuclear Safety (HCTISN). This new authority aims to inform, consult and debate 

about the risks relevant to nuclear activities and their impacts on people’s health, environment and nuclear safety. Andra 

contributes to the functioning of CLI in disposal facilities at Manche and Aube Departments, and CLIS of the 

underground Laboratory at Meuse and Haute-Marne departments. This paper will present these contributions and how 

Andra’s action helps to reach the goals of information and exchange with the people around its facilities. Concerning the 

local economic development, there are specific organizations or schemes, depending on the facility.: - Local taxes 

contributions based on the disposal facilities activities as is usual in France. - A High Level Committee (CHN) and two 

public interest groups (GIP) in Meuse and Haute-Marne departments have been set up since 1991 and 2005. Andra is 

represented in these three institutions, but they are not funded at all by Andra. This paper will show the Andra’s 

involvement in the local economic and territorial developments. Within this general framework Andra has developed 

information and exchanges actions with the stakeholders and the inhabitants around its facilities. Examples of these 

actions will be presented also. (*) -Planning Act No. 2006-739 of 28 June 2006 concerning the sustainable Management 

of Radioactive materials an waste. - Act No. 2006-686 of 13 June 2006 on Transparency and Security in the Nuclear 

Field. 

4)� 40256 – Public Involvement in the Closure of Fernald 

John Bradburne, Bradburne Consulting LLC (USA) 

<strong>The</strong> closure of Fernald set the standard for public involvement in reaching an environmentally acceptable end state 

for DOE sites. All stakeholders including the client (DOE), workforce, residents, state regulators and Federal regulators 

were engaged in determining the path to closure and what endstate at closure would be acceptable. 

SESSION L8: L/ILW Poster 

1) 40006 – A GoldSim Modeling Approach to Safety Assessment of an LLW Repository System 

Youn Myoung Lee, Jongtae Jeong, Jongwon Choi, KAERI (Korea Rep.) 

An assessment program for the safety assessment and performance evaluation of a low- and intermediate level 

waste (LILW) repository system has been developed by utilizing GoldSim, by which nuclide transports in the near- and 

far-field of a repository as well as a transport through a biosphere under various natural and manmade disruptive events 

affecting a nuclide release could be modeled and evaluated. To demonstrate its usability, some illustrative cases under the 

selected scenarios including the influence of degradation of man-made barriers and the possible disruptive events caused 

by an accidental human intrusion or an earthquake have been investigated and illustrated for a hypothetical LILW 

repository. Even though all the parameter values applied to a hypothetical repository are assumed without any real base, 

the illustrative cases are very informative especially when seeing the result of the probabilistic calculation with scenarios 

which are possibly happen for nuclide release and further transport in and around the repository system. 

2) 40011 – Gas Migration Mechanism of Saturated Highly-compacted Bentonite and its Modeling 

Yukihisa Tanaka, Michihiko Hironaga, Koji Kudo, CRIEPI (Japan) 

In the current concept of repository for radioactive waste disposal, compacted bentonite will be used as an 

engineered barrier mainly for inhibiting migration of radioactive nuclides. Hydrogen gas can be generated inside the 

engineered barrier by anaerobic corrosion of metals used for containers, etc. If the gas generation rate exceeds the 

diffusion rate of gas molecules inside of the engineered barrier, gas will accumulate in the void space inside of the 

engineered barrier until its pressure becomes large enough for it to enter the bentonite as a discrete gaseous phase. It is 

expected to be not easy for gas to entering into the bentonite as a discrete gaseous phase because the pore of compacted 

bentonite is so minute. <strong>The</strong>refore the gas migration tests are conducted in this study to investigate the mechanism of gas 

migration. On the basis of the experimental facts obtained through the gas migration tests, possible gas migration 

mechanism is proposed. A simplified method for calculating gas pressure at large breakthrough, which is defined as a 

sudden and sharp increase in gas flow rate out of the specimen is also proposed. 

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3) 40012 – Development of numerical simulation method for gas migration through highly-compacted bentonite 

using model of two-phase flow through deformable porous media 

Yukihisa Tanaka, CRIEPI (Japan) 

In the current concept of repository for radioactive waste disposal, compacted bentonite will be used as an 

engineered barrier mainly for inhibiting migration of radioactive nuclides. Hydrogen gas can be generated inside of the 

engineered barrier by anaerobic corrosion of metals used for containers, etc. It is expected to be not easy for gas to 

entering into the bentonite as a discrete gaseous phase because the pore of compacted bentonite is so minute. <strong>The</strong>refore it 

is necessary to investigate the effect of gas pressure generation and gas migration on the engineered barrier, peripheral 

facilities and ground. In this study, a method for simulating gas migration through the compacted bentonite is proposed. 

<strong>The</strong> proposed method can analyze coupled hydrological-mechanical processes using the model of two-phase flow 

through deformable porous media. Validity of the proposed analytical method is examined by comparing gas migration 

test results with the calculated results, which revealed that the proposed method can simulate gas migration behavior 

through compacted bentonite with accuracy. 

4) 40020 – Planning of Large-scale In-situ Gas Generation Experiment in Korean Radioactive Waste Repository 

Juyoul Kim, Sukhoon Kim, FNC Technology Co.(Korea Rep.); 

Jinbeak Park, Sungjoung Lee, KRMC (Korea Rep.) 

In the Korean LILW (Low- and Intermediate-Level radioactive Waste) repository at the Gyeongju site, the 

degradation of organic wastes and the corrosion of metallic wastes and steel containers would be important processes that 

a?ect repository geochemistry, speciation and transport of radionuclides during the lifetime of a radioactive waste 

disposal facility. 

Gas is generated in association with these processes and has the potential to pressurize the repository, which can 

promote the transport of groundwater and gas, and consequently radionuclide transport. Microbial activity plays an 

important role in organic degradation, corrosion and gas generation through the mediation of reduction–oxidation 

reactions. 

<strong>The</strong> Korean research project on gas generation will be performed by Korea Radioactive Waste Management 

Corporation (KRMC). A large-scale in-situ experiment will form a central part of the project, where gas generation in 

real radioactive low-level maintenance waste from nuclear power plants will be studied. 

In order to examine gas generation from an LILW repository which is being constructed at present by 2012, two 

large-scale facilities for the gas generation experiment (GGE) will be established, each equipped with a concrete 

container for 16 or 9 drums of LILW from Korean nuclear power plants. Each container will be enclosed within a 

gas-tight and acid-proof steel tank. <strong>The</strong> experiment facility will be filled with ground water that provides representative 

chemical conditions and microbial inoculation in the repository near field. 

In the experiment, the design includes long-term monitoring for the rate and composition of gas generated, and 

aqueous geochemistry and microbe populations present at various locations through on-line analyzers and manual 

sampling. 

A main schedule for establishing the experiment facility is as follows: Completion of the detailed design until the 

second quarter of the year 2010; Completion of the manufacture and on-site installation until the second quarter of the 

year 2011; Start of the operation and monitoring from the third quarter of the year 2011. 

5) 40024 – Estimation and measurement of porosity change in cement paste 

Eunyong Lee, Haeryojng Jung, Ki-jung Kwon, KRMC (Korea Rep.); 

Do-Gyeum Kim, Korea Institute of Construction Technology (Korea Rep.) 

Laboratory-scale experiments were performed to understand the porosity change of cement pastes. <strong>The</strong> cement 

pastes were prepared using commercially available Type-I ordinary Portland cement (OPC). As the cement pastes were 

exposed in water, the porosity of the cement pastes sharply increased; however, the slow decrease of porosity was 

observed as the dissolution period was extended more than 50 days. As expected, the dissolution reaction was 

significantly influenced by w/c raito and the ionic strength of solution. A thermodynamic model was applied to simulate 

the porosity change of the cement pastes. It was highly influenced by the depth of the cement pastes. <strong>The</strong>re was porosity 

increase on the surface of the cement pastes due to dissolution of hydration products, such as portlandite, ettringite, and 

CSH. However, the decrease of porosity was estimated inside the cement pastes due to the precipitation of cement 

minerals. 

6) 40082 – Separation and Recovery of Sodium Nitrate from Low-level Radioactive Liquid Waste by 

Electrodialysis 

Yoshihiro Meguro, A Kato, Y Watanabe, Kuniaki Takahashi, JAEA (Japan) 

Low-level radioactive liquid wastes including TRU elements have been generated from several treatments in 

reprocessing facilities. <strong>The</strong>se wastes usually include highly concentrated sodium nitrate. In Japan the liquid wastes are 

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planned to dispose in geological layer with HLW following solidification of them and in this plan it is considered that 

leached nitrate from the solidified materials and its decomposition products would change sorption coefficients of 

radionuclides and then their migration in the long-term would be influenced. <strong>The</strong>refore a method to remove the nitrate 

ions prior to the solidification is required. <strong>The</strong> authors have been developed a separation and recovery method of sodium 

nitrate from the low-level radioactive liquid waste by using an electrodialysis method not only to remove the nitrate but 

also to reduce waste volume and to reuse resources. Here, nitrate ion is recovered as nitric acid and sodium ion as sodium 

hydroxide through ion-exchange membranes. Especially, a ceramic cation-exchange membrane, in which sodium ion was 

selectively transported, was employed to prevent contamination of radioactive nuclides in the recovered sodium 

hydroxide. And also an anion-exchange membrane, in which monovalent anions were selectively transported anions, was 

employed to improve the nitric acid purity. In the present paper, we showed our recent results of investigation for 

transport efficiency of the target ions through the membranes and their selectivity. Aqueous solutions of 5 mol/L sodium 

nitrate containing a small amount of potassium, cesium, and strontium ions or containing nitrite, carbonate, sulfate, and 

phosphate ions were prepared as simulated waste solutions. As a sodium-selective cation-exchange membrane, 

NaSICON membrane produced by Ceramatec Inc. was employed and NEOSEPTA ACS membrane produced by 

ASTOM Co. as a mono-anion permselective membrane. Sodium ions were transferred through the NaSICON membrane 

as much as the quantity of electricity used in the electrodialysis, but only hundredth to 10thousandth amount of potassium, 

cesium, and strontium were transferred through the membrane. <strong>The</strong> sodium ions could be effectively and selectively 

removed from the highly concentrated solution using the NaSICON membrane. Through the ACS membrane monovalent 

anions were efficiently transferred, but less than 10% of multivalent anions were moved at which 90% nitrate was. It was 

confirmed from these results that the electrodialysis method with the ion-selective membranes was useful procedure to 

remove and recover sodium nitrate from the liquid waste. This work was funded by ANRE: Agency for Natural 

Resources and Energy, of METI: Ministry of Economy, Trade and Industry, of Japan. 

7) 40109 – Study on Mechanical Influence of Gas Generation and Migration on Engineered Barrier System in 

adioactive Waste Disposal Facility 

Mamoru Kumagai, JNFL (Japan); Shuichi Yamamoto, Kunifumi Takeuchi, Obayashi Corporation (Japan); 

Yukihisa Tanaka, Michihiko Hironaga, CRIEPI (Japan) 

In Japan, some radioactive waste with a relatively higher radioactivity concentration from nuclear facilities is to be 

packaged in rectangle steel containers and disposed of in sub-surface disposal facilities, where normal human intrusion 

rarely occurs. After the closure of a facility, its pore is saturated with groundwater. If the dissolved oxygen of the pore 

water is consumed by steel corrosion, hydrogen gas will be generated from the metallic waste, steel containers, and 

reinforcing bars of concrete mainly by anaerobic corrosion. If the generated gas accumulates and the gas pressure 

increases excessively in the facility, the facility’s barrier performance might be degraded by mechanical influences such 

as crack formation in cementitious material or deformation of bentonite material. 

Firstly, in this study, we assessed the time evolution of the gas pressure and the water saturation in a sub-surface 

disposal facility by using a multi-phase flow numerical analysis code, GETFLOWS, in which a pathway dilation model is 

introduced and modified in order to reproduce the gas migration mechanism through the highly compacted bentonite. 

Next, we calculated the stress applied to the engineered barriers of the facility from the results of the time evolution 

of the pressure and the saturation. <strong>The</strong>n, we conducted a mechanical stability analysis of the engineered barriers by using 

a nonlinear finite element code, ABAQUS, in order to evaluate their performances after the closure of the facility. 

8) 40192 – Liquid Radioactive Wastes in the Republican Point of Burial Radioactive Waste of Tajikistan 

Nazirzhon Buriev, Physical-Technical Institute of the Academy of Sciences (Tajikistan) 

Liquid radioactive wastes (LRW), as different from the solid ones, have unstable structure and occupy fairly big 

spaces, requiring serious attention to be paid to them and relatively big expenses for arrangement of their storage, 

transportation and disposal. Because of their unstable nature, the LRW constitute great hazards for environment and 

human ecosystem. Under implementation of the <strong>International</strong> Scientific and Technical Center (ISTC) Project ?1082.3 « 

Inspection of Republican Point of Burial Radioactive Wastage with the Purpose of Development of the Technical 

Decision on its Rehabilitation », in accordance with the work plan, there has been projected selection of samples from the 

container for storage of liquid radioactive wastes (LRW), their radiometric and spectral studies, and making decisions for 

LRW further treatment and disposal. Preliminary surveys of the container for storage of LRW revealed that the volume 

of the container is 200m3. LRW in the container were brought and received for disposals from 1967 till 1974 years. At 

present time, the volume of LRW in the container is around 120m3. In accordance with account documents, the general 

activity of LRW in the container makes approximately 0,5 Cu, but these data may be not correct. Activity Indices of 

LRW Stored in the Container Index Specific Activity (Bk/l) Total Activity (Bk) Withdrawal Level (Bk) ? ? 30 3.6 ? 106 

- ? ? 2100 25.2 ? 107 - 137Cs 670 804 ? 105 1 ? 104 226Ra 30 3.6 ? 106 1 ? 104 90Sr 1700 2.0 ? 107 1 ? 104 60?? 5.2 

6.24 ? 105 1 ? 105 3? 2.5 ? 103 3.0 ? 108 1 ? 109 40? 1.6 1.92 ? 105 1 ? 106 

<strong>The</strong> tabled data allow to making the following conclusions: 1. During the long storage of LRW, there turned to 

occur gravitational stratification in the container, with the LRW being stratified for liquid, bottom gels and solid 

sedimentation; bottom sedimentation contain both, big particles (more than 0.5 mm) and small particles (less than 0.5 

mm). 2. Radioactive substances may be in the form of water-soluble ions, gels and solid particles. Gas layer over the 

liquid may have considerable content of radon. 3. As to specific activity, the liquid may be considered as a low-active 

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radioactive substance. At the same time, gels and bottom sedimentation may have high levels of radioactivity. Under the 

project implementation, there has been developed a contact-free suction way of sampling from the container for storage 

of LRW. <strong>The</strong>re has been designed a unit for taking samples from different depths in the container and placing them into a 

special vessel for their storage and transportation. Results of analyses of physical, chemical, radiometric and spectral 

features of the selected samples will allow to development of a technology for the container content decontamination and 

transformation of radioactivity of LRW stored in the container into a stable and compact form for their further disposal. 

9) 40197 – Inspection of Burial Points of Radioactive Waste in the Republic of Tajikistan 

Nazirzhon Buriev,Physical-Technical Institute of the Academy of Sciences (Tajikistan); 

Akram Juraev, Dzhamshed Abdushukurov, 

S.U.Umarov Physical Technical Institute of Academy of Sciences of Republic of Tajikistan, (Tajikistan) 

Under implementation of the <strong>International</strong> Scientific and Technical Center (ISTC) Project ?1082.3 « Inspection of 

Republican Point of Burial Radioactive Wastage with the Purpose of Development of the Technical Decision on its 

Rehabilitation », there has been made a survey of the radioactive waste disposal sites in the RPBRW area for the whole 

period of its activity, starting from 1962 and up to the present time. Before commencement of practical surveys, we had 

studied the following: the RPBRW documentation, foundation history, geographical location, seismic and climatic 

conditions of the area, engineering and technical documentation, history of previous radioactive waste disposals and the 

present state of affairs. <strong>The</strong>re had been arranged meetings and interviews with the RPBRW officers and workers, during 

which information of interest was collected. <strong>The</strong>re had been made a gamma- and neutron-survey of all the RPBRW 

facilities located in the Center’s restricted, controlled and observation zones. Results of the surveys are provided in the 

form of figures put on the facility routing survey maps. During the surveys, there had been selected samples of soils, silts, 

vegetation, sewage waters, atmospheric precipitation and tap water throughout the RPBRW area. In particular, there had 

been selected: in the restricted zone – 109 samples, in the controlled zone – 13 samples, in the observation zone – 91 

samples, and in the adjacent territory – 22 samples. In total, there were taken 235 samples out of which: soils – 210 

samples, vegetation – 10 samples, silts – 2 samples, sewage waters – 10 samples, tap water – 2 samples, and snow – 1 

sample. All the samples were selected, packed, marked, and transported to laboratory to prepare them for radiometric and 

spectral studies in strict observance of the methodical guidelines developed. <strong>The</strong> tests were made by means of HPG 

detector manufactured by a firm “Canberra”. In the restricted zone, soil samples were taken from the surface as well as 

from pits, to trace soil migration of radionuclides. Pits were dug out up to 2m deep around radioactive waste disposal 

sites down the slope of a hill. <strong>The</strong>re are no open water sources in the RPBRW area. We selected precipitation waters 

flowing across the RPBRW area, which may bear important information on radiation situation. Samples were taken in the 

spring time right after raining. Selection points were atmospheric precipitation drains and natural collectors near the 

RPBRW facilities. On the same sites there were taken the samples of silt borne by waste waters. Vegetation samples 

were annual grasses and bushes thriving in the RPBRW area, as well as arboraceous samples (acacia tree). To trace 

possible migration of radionuclides into the trunk of a tree growing in the close vicinity of the radioactive waste disposal 

site, there had been taken samples of its bark, trunk, small annual branches and leaves. Results of the sample laboratory 

tests are incorporated into the Project T-1082.3 general database compiled as tables in Microsoft Office Excel format. 

10) 40198 – Pre-treatment Method of Powder Waste for Vitrification 

Deun-Man Kim, Hyun-Je Cho, Jong-Kil Park, Nuclear Engineering & Technology Institute (Korea Rep.) 

Solidification techniques to reduce the volume of the waste have been developed using binder materials such as 

cement, paraffin, and polymer. A vitrifying technique that realizes high volume reduction and stably confines 

radioactivity within the glass material was researched with dried powder waste. Boric acid waste in PWRs is dried on 

Concentrated Waste Dry System (CWDS). In the case of feeding the powder wastes into a cold crucible induction melter 

without pre-treatment, there is a possibility that the powder waste may enter the off-gas treatment system before mixture 

with melted glass and the feeding pipes could be clogged because of accumulation of these waste. A method to pre-treat 

the concentrated waste by pelletizing powder waste was studied for a vitrification process. Final products were satisfied 

with the requirements of hardness and friability. This research presents the pretreatment method of powder waste to 

apply on vitrification facilities. 

11) 40208 – Characterization and Pre-treatment of LLW in Turkey 

Ahmet Erdal Osmanlioglu, Turkish Atomic Energy Authority (Turkey) 

Pre-treatment of radioactive waste is the first step in waste management program that occurs after waste generation 

from various applications in Turkey. Pre-treatment and characterization practices are carried out in Radioactive Waste 

Management Unit (RWMU) at Cekmece Nuclear Research and Training Center (CNRTC) in Istanbul. This facility has 

been assigned to take all low-level radioactive wastes generated by nuclear applications in Turkey. <strong>The</strong> wastes are 

generated from research and nuclear applications mainly in medicine, biology, agriculture, quality control in metal 

processing and construction industries. <strong>The</strong>se wastes are classified as low- level radioactive wastes. Pre-treatment 

practices cover several steps. In this paper, main steps of pre-treatment and characterization are presented. Basically these 

are; collection, segregation, chemical adjustment, size reduction and decontamination operations. 

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12) 40221 – Decontamination of Radioactive Concrete Wastes by <strong>The</strong>rmal and Mechanical Processes 

B.Y. Min, Ki Won Lee, Un Soo Chung, W.K Choi, Kune-Woo Lee, KAERI (Korea Rep.) 

<strong>The</strong> purpose of this paper is to provide the results of a series of volume reduction tests consucted in thermal and 

mechanical processes. Korea Atomic Energy Research Institites (KAERI) has developed volume reduction technology 

applocable to an activated heavy concrete waste generated by dismantled Korea research reactor 2 (KRR-2) and a 

uranium conversion plant (UCP). the volume reductuin could be achieved above 70% by lab scale test by thermal and 

mechanical separation processes. Pilot test were performed with radioactively contaninated dismantled concrete waste. 

<strong>The</strong> developed processes were quite effective for volume reduction of radioactively contaminated dismantled concrete 

waste. 

13) 40302 – Latex Particles Functionalized with Transition Metals Ferrocyanides for Cesium Uptake and 

Decontamination of Solid Bulk Materials 

Dmitry Marinin, Valentin Avramenko, Dmitry Marinin, Valentin Sergienko, 

Institute of Chemistry FEDRAS (Russia); 

Veniamin Zheleznov, Irina Sheveleva, Russian Academy of Sciences (Russia) 

Decontamination of spent ion-exchange resins, corrosion-unstable metal structures, soil, ground, and construction 

materials contaminated by fission, corrosion and transuranic radionuclides remains one of the most urgent and 

complicated ecological problems. Among the existing methods having different efficiencies in regard to such materials 

decontamination, application of selective sorbents put into a humid medium to be decontaminated (ground, bulk 

materials) appears to be rather extensive. However, the efficiency of such an approach is significantly limited by 

difficulties concerned with uniform sorbent distribution in porous media and completeness of spent sorbents removal for 

final disposal. In this paper we suggest a principally new approach to preparation of colloid-stable selective sorbents for 

cesium uptake using immobilization of transition metals (cobalt, nickel, and copper) ferrocyanides in nanosized 

carboxylic latex emulsions. <strong>The</strong> effects of ferrocyanide composition, pH, and media salinity on the sorption properties of 

the colloid-stable sorbents toward cesium ions were studied in solutions containing up to 200 g/l sodium nitrate or 

potassium chloride. <strong>The</strong> sorption capacities of the colloid sorbents based on mixed potassium/transition metals 

ferrocyanides were in the range 1.45-1.86 mol Cs/mol ferrocyanide with the highest value found for the copper 

ferrocyanide. It was shown that the obtained colloid-stable sorbents were capable to penetrate through bulk materials 

without filtration that makes them applicable for decontamination of solids, e.g. soils, zeolites, spent ion-exchange resins 

contaminated with cesium radionuclides. After decontamination of liquid or solid radioactive wastes the colloid-stable 

sorbents can be easily separated from solutions by precipitation with cationic flocculants providing localization of 

radionuclides in a small volume of the precipitates formed. Besides, functionalized latex particles can be used for 

preparation of carbon fiber/ferrocyanide composite materials for cesium uptake using electrodeposition method. 

Application of the carbon fibers as an inert support for ferrocyanides, in general, significantly improves the sorption 

kinetics, but washing out of ferrocyanide fines from the fiber surface limits the potential of such materials. When 

ferrocyanides are deposited in a form of nanocrystals stabilized by latexes which undergo electropolymerization on the 

fiber surface, the thin polymeric film formed substantially improves the stability of the composite and prevents loss of 

ferrocyanide during sorbent application. <strong>The</strong> effect of electrodeposition conditions on composite morphology, 

ferrocyanide loading and cesium distribution coefficient in media with different salinity has been discussed. 

SESSION H11: SF/TRU/HLW Poster 

1) 40001 – Investigation of Colloid-facilitated Effects on the Radionuclides Migration in the Fractured Rock with 

a Kinetic Solubility-Limited Dissolution Model 

Chun-Ping Jen,Nation Chung Cheng University (Taiwan); 

Neng-Chuan Tien, Energy and Environment Research Laboratories, 

Industrial Technology Research Institute (Taiwan) 

Nuclides can move with the groundwater either as solutes or colloids, where the latter mechanism generally results 

in much shorter traveling time as they interact strongly with solid phases, such as actinides. It is therefore essential to 

assess the relative importance of these two transport mechanisms for different nuclides. <strong>The</strong> relative importance of 

colloids depends on the nature and concentration of colloids in groundwater. Plutonium (Pu), neptunium (Np), uranium 

(U) and americium (Am) are four nuclides of concern for long-term emplacement of nuclear wastes at potential 

repository sites. If attached to iron oxide, clay or silica colloids in groundwater. Strong sorption of the actinides by 

colloids in groundwater may facilitate transport of these nuclides along potential flow paths. Solubility-limited 

dissolution model models can be used to determine the release of the safety assessment for nuclear waste in geological 

disposal sites. <strong>The</strong> present study investigates the effect of colloid on the transport of solubility limited nuclide under the 

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kinetic solubility limited dissolution (KSLD) boundary condition in fractured media. <strong>The</strong> release rate of nuclide would 

proportional to the difference between the saturation concentration and the inlet aqueous concentration of nuclide. <strong>The</strong> 

presence of colloids could decrease the aqueous concentration of nuclide and thus could increase the release flux of 

nuclide from the waste form. 

2) 40013 – Modeling hydraulic conductivity and swelling pressure of several kinds of bentonites affected by 

salinity of water 

Yukihisa Tanaka, Takuma Hasegawa, Kunihiko Nakamura, CRIEPI (Japan) 

In case of construction of repository for radioactive waste near the coastal area, the effect of salinity of water on 

hydraulic conductivity as well as swelling pressure of bentonite as an engineered barrier should be considered because it 

is known that the hydraulic conductivity of bentonite increases and swelling pressure decreases with increasing in salinity 

of water. Though the effect of salinity of water on hydraulic conductivity and swelling pressure of bentonite has been 

investigated experimentally, it is necessary to elucidate and to model the mechanism of the phenomenon because various 

kinds of bentonites may possibly be placed in various salinity of ground water. Thus, in this study, a model for evaluating 

hydraulic conductivity as well as swelling pressure of compacted bentonite is proposed considering the effect of salinity 

of water as follows : 

a) Change in number of flakes of a stack of montmorillonite because of cohesion 

b) Change in viscosity of water in interlayer between flakes of montmorillonite. 

Quantitative evaluation method for hydraulic conductivity and swelling characteristics of several kinds of bentonite 

under saline water is proposed based on the model mentioned above. 

3) 40018 – Current R & D activities of the study on long-term stability of geological environments 

Takahiro Hanamuro, Kenichi Yasue, Yoko Saito-Kokubu,Koichi Asamori, 

Tsuneari Ishimaru, Koji Umeda, JAEA (Japan) 

Japanese islands are located in a tectonically active zone, where earthquakes and volcanic eruptions frequently 

occur. <strong>The</strong>refore the understanding of the long-term stability of geological environment is required for assessing the 

long-term behaviour of the geological disposal system of high level radioactive waste (HLW) in Japan. <strong>The</strong> Japan 

Atomic Energy Agency (JAEA) is promoting the establishment of investigation and assessment methods of the long-term 

stability of geological environment necessary for site selection and safety assessment of HLW geological disposal. 

<strong>The</strong> Nuclear Waste Management Organization of Japan (NUMO) is the implementation body of HLW disposal in 

Japan. <strong>The</strong> preliminary investigation areas for HLW disposal site, selected by NUMO in the future, are supposed to be 

selected excluding the zones affected by already-known active faults identified by nationwide and site-specific literature 

surveys. <strong>The</strong>y are also supposed to be selected with excluding the area within 15 km radius of the center of a Quaternary 

volcano and the zones affected by thermal and hydrothermal activities identified by site-specific literature surveys. For 

uplift, denudation and climatic/sea-level changes, it is necessary that the change of geological environment caused by 

uplift/denudation and climatic/sea-level changes is assessed for HLW geological disposal system. 

For seismicity and faulting, some detection techniques for active faults with no surface expression, by using helium 

isotope ratio of hot spring gas or detection of hydrogen gas, and studying on the assessment of fault activities are 

developed. For volcanism and geothermal activity, heat source of geothermal anomaly area in the non-volcanic region are 

considered and some detection techniques for high-temperature fluid and magma at deep underground, by using seismic 

tomography, magnetotelluric method and helium isotope ratio of hot spring gas, are constructed. For uplift, denudation 

and climatic/sea-level changes, a methodology to understand future topographic change with time is developed. Also, for 

dating techniques as an essential part to proceed on these studies, C-14 dating by using AMS and (U-Th)/He dating by 

using QMS and ICP-MS have developed, and Be-10 dating by using AMS has been being developed. 

We are planning the establishment of assessment methods for long-term stability of geological environment; 

assessment of activities of faults encountered by underground excavation, development of long-term estimation methods 

of volcanisms and hydrothermal activities, and hydrogeological analyses considering topographic change in the future. 

4) 40019 – In Situ Stress Measurements in Siliceous Mudstones at Horonobe Underground Research Laboratory, 

Japan 

Hiroyuki Sanada, Takahiro Nakamura, Yutaka Sugita, JAEA (Japan) 

As part of the research and development program on the HLW geological disposal, JAEA has been implementing 

the Horonobe Underground Research Laboratory (URL) Project investigating sedimentary rock formations distributed in 

Horonobe area, Hokkaido, Japan. To optimize the design and construction of underground excavations for any HLW 

repository, a thorough evaluation of initial stress condition will be required. Studies of initial stress conditions in 

sedimentary rock aren't enough, since important underground structures such as those for underground power plants and 

caverns for storage of liquefied petroleum or natural gas in Japan have mainly been constructed in hard rock. 

Additionally, initial stress measurements in sedimentary rocks are inherently more difficult compared to hard rock. <strong>The</strong> 

authors have been implementing the research and development program at Horonobe to clarify the in situ stress 

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conditions in siliceous mudstones at the URL. <strong>The</strong> objective of this work is to establish a strategy for an in situ stress 

measurement program for geological disposal and to develop an understanding of the in situ stress conditions in the deep 

underground formed by the sedimentary rocks. <strong>The</strong> application of several stress measurement methods to the Horonobe 

siliceous mudstones carried out during the surface-based investigations and the investigations during construction of the 

underground facilities, as well as information on the initial stress state around the Horonobe URL are described in this 

paper. During the surface-based investigations, determination of deep in situ stress was done using hydraulic fracturing 

(HF), borehole breakout information in deep boreholes and core-based methods such as AE and DSCA. Subsurface 

investigations during construction of the underground facilities utilized, the Compact Conical Ended Borehole 

Overcoring (CCBO) method and hydraulic testing of pre-existing fractures (HTPF) were conducted in order to validate 

results from initial stress measurements in the surface-based investigations. HF results indicate that horizontal maximum 

and minimum principal stresses increase linearly with depth. <strong>The</strong> minimum principal stress is almost equivalent to 

overburden pressure. <strong>The</strong> maximum principal stress estimated from the HF and borehole breakout data is almost E-W. 

This is similar to the tectonic movement direction in the vicinity of the Horonobe URL. Due to tectonic movement, 

horizontal maximum stress is almost 1.5 times larger than the horizontal minimum stress. Vertical stress determined from 

HTPF during construction of the underground facilities is almost equal to the overburden pressure. 

5) 40038 – Low alkaline Cement Used in the Construction of a Gallery in the Horonobe Underground Research 

Laboratory 

Masashi Nakayama, Haruo Sato, Yutaka Sugita, Seiji ITO, JAEA (Japan); 

Masashi Minamide, Yoshito Kitagawa, Taisei Cooperation (Japan) 

In Japan, any high level radioactive waste repository is to be constructed at over 300m depth below surface. Tunnel 

support is used for safety during the construction and operation, and shotcrete and concrete lining are used as the tunnel 

support. Concrete is a composite material comprised of aggregate, cement and various additives. Low alkaline cement 

has been developed for the long term stability of the barrier systems whose performance could be negatively affected by 

highly alkaline conditions arising due to cement used in a repository. Japan Atomic Energy Agency (JAEA) has 

developed a low alkaline cement, named as HFSC (High fly-ash silicafume cement), containing over 60wt% of 

silica-fume (SF) and coal ash (FA). JAEA are presently constructing an underground research laboratory (URL) at 

Horonobe for research and development in the geosciences and repository engineering technology. HFSC was used 

experimentally as the shotcrete material in construction of part of the 140m deep gallery in Horonobe URL. <strong>The</strong> 

objective of this experiment was to assess the performance of HFSC shotcrete in terms of mechanics, workability, 

durability, and so on. HFSC used in this experiment is composed of 40wt% OPC (Ordinary Portland Cement), 20wt% SF, 

and 40wt% FA. This composition was determined based on mechanical testing of various mixes of the above 

components. Because of the low OPC content, the strength of HFSC tends to be lower than that of OPC in normal 

concrete. <strong>The</strong> total length of tunnel using HFSC shotcrete is about 73m and about 500m3 of HFSC was used. This 

experimental construction confirmed the workability of HFSC shotcrete. Although several in-situ experiments using low 

alkaline cement as shotcrete have been performed at a small scale, this application of HFSC at the Horonobe URL is the 

first full scale application of low alkaline cement in the construction of a URL in the world. In the paper, we present 

detailed results of the in-situ construction test and the future works. 

6) 40040 – Effects of Nitrate on Nuclide Solubility for Co-location Disposal of TRU Waste and HLW 

Gento Kamei, Morihiro Mihara, JAEA (Japan); Toshiyuki Nakazawa, 

Norikazu Yamada, Mitsubishi Materials Corp. (Japan) 

TRU wastes are generated by reprocessing spent fuel from nuclear power plants and by fabricating MOX fuel in 

Japan. Some of the TRU wastes are expected to be disposed of deep underground to isolate it from the biosphere in the 

long-term. To optimize the disposal of TRU waste, a co-location disposal with high level waste, HLW, is being 

considered. A part of TRU waste includes a large amount of nitrate salt, the effects of which have to be evaluated in a 

safety assessment of co-location disposal. 

Solubility is one of the important parameters for the safety assessment of HLW disposal. Large concentrations of 

nitrate ions from TRU waste might affect the oxidation state and consequently the solubility of different radionuclides in 

the HLW waste. In addition, it is necessary to consider complex formation of nitrate ions with radionuclides, as well as 

the formation of ammonia by microbes and/or by reactions with reducing materials in the disposal facility. Consequently, 

complex formation of ammonia with radionuclides must also be evaluated. 

In the current study, the effects of nitrate salt on radionuclide solubility were investigated experimentally with 

consideration given to the above perturbations. Solubility experiments of important and redox sensitive radionuclides, 

Tc(IV), Np(IV) and Se(0), were performed using various concentrations of sodium nitrate (NaNO3) and of Np(V) in 

NaNO3 solutions to investigate complex formation with NO3- ions. Solubility experiments of Pd(II), Sn(IV) and Nb(V) 

using ammonium chloride (NH4Cl) solution were also undertaken to investigate complex formation with NH3/NH4+ 

ions. A chemical equilibrium model was applied to assist the interpretation of the experimental results. No significant 

solubility enhancement was observed for Np and Se due to oxidation by nitrate ions. Tc solubility was, however, 

increased by one order of magnitude under high nitrate concentrations. Solubility enhancement by complex formation of 

nitrate ions with Np(V) was not observed. Solubility enhancement by complex formation of Sn and Nb were not also 

observed, only Pd solubility was increased by complex formation with NH3/NH4+ ions. Tendencies of the enhancement 

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of Pd solubility were explained by the chemical equilibrium model. 

This work was funded by ANRE: Agency for Natural Resources and Energy, of METI: Ministry of Economy, Trade 

and Industry, of Japan. 

7) 40047 – A study on groundwater infiltration in the Horonobe area, northern Hokkaido, Japan 

Hideharu Yokota, Yamamoto Yoichi, Keisuke Maekawa, Minoru Hara, JAEA (Japan) 

It is important for assessing the safety of geological disposal of high-level radioactive waste to understand 

groundwater flow as the driving force of mass transport. In the groundwater-flow simulation, hydraulic boundary 

conditions are required, including groundwater-recharge rates. In the Horonobe area of northern Hokkaido, the Japan 

Atomic Energy Agency (JAEA) has been carrying out the Horonobe Underground Research Laboratory (URL) Project to 

understand characteristics of the geological environment. To obtain various hydrological data to estimate the recharge 

rate by water balance, meteorological observation and observation of river flux, etc. in the Horonobe area (snowy cold 

region) have been carried out. However, infiltration of water from the surface is difficult to clarify in detail because water 

near ground surface is sensitive to external influence such as climatic variations. It is important for precise evaluation of 

groundwater flow to understand shallow groundwater-flow systems (ground surface to tens of meters at depth) as a part 

of hydraulic boundary conditions. In the Horonobe area, subsurface temperature and soil moisture content have been 

observed at the URL (GL-0.7m to GL-2.3m) site since 2005 and at Hokushin Meteorological Station (HMS, GL-0.1m to 

GL-1.1m) since 2008. As results of these observations, it is clear that similar processes operate at both sites. Subsurface 

temperatures become lower with depth in the summer and higher with depth in the winter. <strong>The</strong> lowest subsurface 

temperatures at the shallowest and deepest at depth are observed in the middle of April and early May respectively. 

Concurrently, soil moisture content increases rapidly. In addition, the observed data also show that the subsurface 

temperature is higher than 0°C throughout the year, and keeps decreasing until early May (snow-melting season). From 

these results, it is suggested that, regardless of the air temperature, water at 0°C is supplied from the bottom of 

snow-cover to ground surface by bottom snow melting due to the insulating effect of snow. <strong>The</strong>refore, the subsurface 

temperature firstly becomes the lowest at the shallowest depth. Subsequently, subsurface temperatures at greater depths 

decrease as the cold water infiltrates to depth with time. For the estimation of boundary conditions in groundwater-flow 

simulation, in this study, the shallow groundwater-flow system has been examined qualitatively on the basis of the 

seasonal variation of the groundwater infiltration. Results have revealed the groundwater recharge occurring in a 

snow-covered region. In the future, it is planed that quantitative assessments will be made by the observed data of the 

weighing lysimeter. 

8) 40051 – Effective Use of Uranium Resources and Dissolution of Recovery Uranium Storage Accumulation by a 

Uranium Multi-recycle System 

Yuzo Yamashita, Yuzo Yamashita, Takeshi Nakamura, Kyushu University (Japan) 

<strong>The</strong> uranium recovered from LWR spent fuels, containing an amount of U-235 comparable to that in the natural 

uranium, can be recycled as uranium fuels for LWRs by re-enrichment using a conventional centrifuge cascade. However, 

the remade fuel is inferior to the original fuel produced from the natural uranium on the burn-up performance, because 

the former includes U-236 as neutron absorber which is yielded in the spent fuel and then enhanced in re-enrichment 

process. <strong>The</strong>refore, a few times recycle of recovered uranium may be available but its successive multi-recycle is not 

recommendable because of successive decline in burn-up resulting from U-236 accumulating in remade fuels. In this 

study, an idea of uranium fuel recycling is proposed which is of the feed-back of recovered uranium to a natural uranium 

enriching cascade. Since this cascade processes a mixture of natural and recovered uranium, the product contains U-236 

enriched but diluted with the U-236-free bulk of natural uranium. In this process, the concentration of U-236 in the fuels 

reaches a balance with increase in recycle times. <strong>The</strong> multi-recycling fuels perform the burn-up degrees in PWRs 

declining but comparing favorably with convetional uranium ones. <strong>The</strong> multi-recycling system not only makes effective 

use of nuclear fuel resources, but also dissolves the problem of accumulating inventory in the storage of recovered 

uranium taking the major volume of spent fuels. 

9) 40053 – Advanced ORIENT Cycle - Progress on Fission Product Separation and Utilization 

Isao Yamagishi, Masaki Ozawa, JAEA (Japan); Hitoshi Mimura, Tohoku Univ.(Japan); 

Shohei Kanamura, Koji Mizuguchi, Toshiba Corporation (Japan) 

Fission reaction of U-235 generates more than 40 elements and 400 nuclides in the spent fuel. Among them, 31 

elements are categorized as rare metals. Typical yields of Pd, Ru, Rh (PGM) and Tc as rare metals will reach around 

11kg, 13kg, 4kg and 3kg, respectively per metric ton of the reference FBR spent fuel (150GWd/t, cooled 5 years). Based 

on a ground swell on enhancements of minimization of radio-ecological impacts and economical expenditures, nuclear 

fuel cycle concept itself is required to be changed. Adv.-ORIENT (Advanced Optimization by Recycling Instructive 

Elements) Cycle strategy was hence drawn up for the minimization of radioactive waste and utilization of 

elements/nuclides in the wastes simultaneously., and has been developed at Japan Atomic Energy Agency. <strong>The</strong> present 

paper deals with the separation process of fission products in the Adv.-ORIENT Cycle, which consists of the 

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Abstracts 

chromatographic separation of heat-generating Cs and Sr, and the electrodeposition of Pd, Ru, Rh and Tc. Highly 

functional inorganic adsorbent (silica gel loaded with ammonium molybdophosphate, AMP-SG) and organic 

microcapsule (alginate (ALG) gel polymer enclosed with crown ether D18C6) were investigated for separation of Cs and 

Sr, respectively, from high-level liquid waste (HLLW). <strong>The</strong> AMP-SG adosorbed more than 99% of Cs from the 

simulated HLLW selectively. <strong>The</strong> ALG microcapsule adsorbed 0.0249 mmol/g of Sr and exhibited the order of its 

selectivity: Ba > Sr > Pd >> Ru > Rb > Ag. Separated Cs and Sr will be immobilized in ceramic forms using zeolites for 

the utilization as heat and/or radiation source. <strong>The</strong> electrodeposition is advantageous for both recovery and utilization of 

PGM and Tc because these elements are recovered as metal (Ru, Rh, Pd, (Tc)) and/or oxide (Tc) form on a Pt electrode. 

In the presence of Pd or Rh the reduction of Ru and Tc was accelerated in hydrochloric acid media. This co-deposition 

effect, however, was less in nitric acid. In the simulated HLLW, the redox reaction of Fe(III)/Fe(II) disturbed deposition 

of elements except for Pd. <strong>The</strong> deposits on Pt electrode showed higher catalytic reactivity on electrolytic hydrogen 

production than the original Pt electrode. 

Keywords; Nuclear waste, Separation, Utilization, Adsorption, electrolysis, Cesium, Strontium, Ruthenium, 

Rhodium, Palladium, Technetium, PGM 

10) 40064 – Hydrogeological Characterization based on Long Term Groundwater Pressure Monitoring 

Shuji Daimaru, Ryuji Takeuchi, Masaki Takeda, Masayuki Ishibashi, JAEA (Japan) 

<strong>The</strong> Mizunami Underground Research Laboratory (MIU) is now under construction by the Japan Atomic Energy 

Agency in the Tono area of central Japan. <strong>The</strong> MIU project is being implemented in three overlapping Phases: 

Surface-based Investigation (Phase I), Construction (Phase II) and Operation (Phase III). <strong>The</strong> changes of groundwater 

pressure due to shaft excavation can be considered analogous to a large-scale pumping test. <strong>The</strong>refore, there is the 

possibility that the site scale groundwater field (several km square) can be approximated by the long-term groundwater 

pressure monitoring data from Phase II. Based on the monitoring observations, hydrogeological characteristics were 

estimated using the s–log(t/r^2) plot based on the Cooper-Jacob straight line method. Results of the s-log(t/r^2) plots are 

as follows. <strong>The</strong> groundwater flow field around the MIU construction site is separated into domains by an impermeable 

fault. In other words, the fault is a hydraulic barrier. Hydraulic conductivity calculated from s-log(t/r^2) plots are in the 

order of 1.0E-7(m/s). <strong>The</strong> above results from the long term monitoring during PhaseII are a verification of the 

hydrogeological characteristics determined in the Phase I investigations. Keywords: Large scale pumping test, long term 

pressure monitoring, shaft excavation, s-log(t/r^2) plot, hydrogeological characteristics. 

11) 40065 – Development Of <strong>The</strong> Quality Management System For Borehole Investigations: (2) Quality 

Management Systems For Hydrochemical Investigations, 

Takanori Kunimaru, Kunio Ota, Kenji Amano, JAEA (Japan); 


An appropriate Quality Management System (QMS), which is among the first tools required for repository site 

characterisation, will save on effort by reducing errors and the requirement to resample and reanalyse – but this can only 

be guaranteed by continuously assessing if the system is truly fit-for-purpose and amending it as necessary based on the 

practical experience of the end-users on-site. As part of the national research and development programme for deep 

geological disposal of radioactive waste, the Japan Atomic Energy Agency has established two Underground Research 

Laboratories (URL) based around Horonobe, northern Japan, and Mizunami, central Japan. <strong>The</strong> main aim of the URL 

programme is to define comprehensive techniques for future repository site characterisation. At the Horonobe URL, 

investigation of the geological environment within sedimentary host formations is currently ongoing and one facet of this 

is one of the few examples worldwide (and the first use in Japan) of the study of rock matrix porewater hydrochemistry, 

in conjunction with groundwater hydrochemistry. <strong>The</strong> data and interpretation are described in detail elsewhere but, 

before detailed data interpretation, the first priority should always be an analysis of the data quality and this is addressed 

here. <strong>The</strong> quality of the porewater data have been categorised based on an existing system of ranking groundwater data 

quality, which has been developed for the fractured crystalline rocks of the Fennoscandian Shield. <strong>The</strong> ranks range from 

Category 1 (highest quality) to Category 5 (lowest quality) and of particular importance in the categorisation criteria are 

(i) the degree of porewater contamination with drilling fluid, (ii) indications of sample perturbations such as oxidation or 

CO2 reaction and (iii) the completeness of major ion and isotope analytical data. With the integration of all available 

information on hydrochemistry, hydrogeology and borehole history, more than 150 porewater and 24 groundwater 

datasets have been examined to date. <strong>The</strong> results of the QA audit by multi-operators have been compared in order to 

assess the objectivity of the categorisation, methodology and this clearly shows operator-dependent differences in 

categorisation of the same data set. As such, the initial guidelines for assigning the QA categories have been improved so 

as to reduce dependence on expert judgement and this approach will be discussed in detail here. This approach will 

increase transparency in the data handling and so increase stakeholders’ confidence in both the data set itself and in the 

results obtained using such data for a repository site characterisation. 

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12) 40067 – An Analytical Model on the Sealing Performance of Space for the Design of Buffer Material and 

Backfill Material 

Haruo Sato, JAEA (Japan) 

<strong>The</strong> self-sealing function of spaces between buffer material and overpack, tunnel wall and disposal pit by swelling 

occurred with water penetration is expected for bentonite which will be used as buffer material and part of the backfill 

material in the geological of high-level radioactive waste in Japan. <strong>The</strong> clearance filling properties of Na-bentonite for 

buffer material and backfill material specifications have been studied for distilled water and saline water conditions, for 

example, it is reported that Na-bentonite seals clearance even under saline water conditions in a range of effective 

bentonite densities which are higher than 1.3 kg/dm3, for a bentonite dry density of 1.8 kg/dm3 and a clearance ratio 

(volumetric ratio of clearance volume to total volume including pore and the clearance) of 10 % (1.11 in volumetric 

swelling ratio: Rvs) in experiments for Kunigel-V1 (Na-montmorillonite content ca. 50 wt.%). Although such 

information is useful for judging whether clearance is filled, the filling properties of bentonite depend on groundwater 

condition, silica sand content, montmorillonite content in the bentonite and the bentonite dry density, even though at the 

same effective bentonite density. In the present study, the author constructed an analytical model on the clearance filling 

performance for the design of buffer material and backfill material, based on the swelling properties of 

Na-montmorillonite which is the clay mineral constituent of bentonite. In the modelling, experimental data on the Rvs of 

bentonite (Kunigel-V1) reported so far were analyzed. Consequently, it was found that dry density of the bentonite when 

the free swelling reached equilibrium state was approximately in the range of 0.204-0.241 kg/dm3 for distilled water 

condition. <strong>The</strong>se dry densities are equivalent to the range of 0.106-0.126 kg/dm3 in montmorillonite partial density. On 

the other hand, the Rvs of bentonite decreased for saline water condition. Based on these experimental data for Rvs, the 

author derived an analytical expression to be able to calculate Rvs values against various dry densities and bentonites 

(different montmorillonite and silica sand contents), and showed the manner for judging whether clearance is sealed for 

an arbitrary clearance ratio. In the paper, the author shows the calculated results of Rvs versus dry density under various 

conditions. 

13) 40069 – Current Status of Horonobe URL Project in Construction Phase 

Hironobu Abe, Koichiro Hatanaka, JAEA (Japan) 

Horonobe URL project has been pursued by JAEA (Japan Atomic Energy Agency) to establish and demonstrate site 

characterization methodologies, engineering technologies, and safety assessment methodologies for HLW geological 

disposal in relevant geological environment with sedimentary rock and saline groundwater distributing in Horonobe area, 

Hokkaido, Japan. In the Horonobe URL project, surface-based investigation phase (Phase I) has already completed in the 

year 2005, and then construction phase (Phase II) has initiated in the same year. Currently, construction of the 

underground facilities such as shafts/drifts which were designed in Phase I, investigations of the geological environment 

in the excavated shafts/drifts and confirmation of applicability of engineering technologies has been alternately carried 

out as Phase II activities of the project. At the end of January, 2010, ventilation shaft has reached at the depth of 

GL-250m and east access shaft has reached at the depth of GL210m. Concerning horizontal drifts, construction of 

GL-140m gallery with total length of 184m has completed and GL-250m gallery has partially constructed with the length 

of 42m. During the construction so far, monitoring for the construction safety such as convergence measurements, tunnel 

wall observation, sampling of groundwater and rock, investigations for evaluating excavation damaged zone along 

shaft/drift were carried out. In addition, shotcrete construction test and grout injection test by using low alkaline cement 

material were carried in the horizontal drifts. In this paper, status of the URL construction and research activities 

mentioned above are outlined as the current achievement of the Horonobe URL project. 

14) 40074 – Development of New Ultrafiltration Techniques Maintaining In-Situ Hydrochemical Conditions for 

Colloidal Study 

Daisuke Aosai, Yuhei Yamamoto, Takashi Mizuno, JAEA (Japan) 

Chemical state of elements in groundwater is one of the most important information for understanding behavior of 

elements in underground environment. Chemical state of elements controlled mainly by groundwater physico-chemical 

parameters. Because the change of physico-chemical parameters of groundwater, due to pressure release and oxidation 

during sampling, causes changes in chemical state of elements, systematic methodologies for understanding in situ 

chemical state is required. In this study, in order to understand chemical state of elements in groundwater, an 

ultrafiltration instrument for maintaining in-situ pressure and anaerobic conditions was developed. <strong>The</strong> instrument 

developed in this study for ultrafiltration made of passivated Stainless Used Steel (SUS) materials, was designed to keep 

groundwater samples maintaining in-situ pressure/anaerobic conditions. Ultrafiltration of groundwater was conducted at 

a borehole drilled from the 200 mbGL (meters below ground level) Sub-stage at a depth of 200 m at the Mizunami 

Underground Research Laboratory. Chemical analyses of groundwater were also conducted using samples filtered under 

both pressurized/anaerobic and atmospheric conditions and passivated SUS materials with different elapsed times after 

passivation. <strong>The</strong> results indicate that our ultrafiltration method is suitable for collection of filtered groundwater and 

passivation is an essential treatment before ultrafiltration. Keywords: groundwater, ultrafiltration. 

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Abstracts 

15) 40089 – Sorption Behavior of Iodine on Calcium Silicate Hydrates Formed as a Secondary Mineral 

Keisuke Shirai, Yuichi Niibori, Akira Kirishima, Hitoshi Mimura, Tohoku Univ. (Japan) 

This study examined the sorption behaviors of iodine into CSH gel without dried processes, considering the 

repository system saturated with groundwater after the backfilling. In glove box saturated with N2 gas, each sample of 

CSH gel was synthesized with CaO, SiO2, and distilled water with liquid/solid ratio 20. <strong>The</strong>n, 1 mM iodine solution is 

added into the aqueous solution including the CSH gel with various Ca/Si molar ratios under the isothermal condition 

(298 K). 

In the results, even if the Ca/Si ratio is relatively small (

Abstracts 

determined concentration of Sn-126 was equivalent to 60% of that obtained by ORIGEN2 calculation, which suggests 

that a part of Sn initially existed in the sample UO2 pellet precipitated as insoluble residue during the dissolution process. 

18) 40103 – Selective Uptake of Palladium from High-Level Liquid Wastes by Hybrid Microcapsules 

Hitoshi Mimura, Takashi Sakakibara, Yan Wu, Yuichi Niibori, Tohoku Univ. (Japan); 

Shin-ichi Koyama, Takashi Ohnishi, JAEA (Japan) 

Fine crystalline powders of KCuFC were immobilized with alginate gel polymers by sol-gel methods. <strong>The</strong> uptake 

properties of KCuFC-microcapsules (KCuFC-MC) were examined by batch and column methods. <strong>The</strong> size of 

KCuFC-MC particle was estimated to be about 1 mm in diameter, and KCuFC powders were uniformly dispersed in 

KCuFC-MC particles. <strong>The</strong> uptake rate of Pd2+ for KCuFC-MC was attained within 3 d, and the uptake of Pd2+ was 

found to be independent of the temperature and coexisting HNO3 concentration. As for the breakthrough properties of 

Pd2+ through a column packed with KCuFC-MC, a breakpoint of 5% breakthrough was enhanced with lowering of flow 

rate and independent of coexisting HNO3 concentration. <strong>The</strong> Pd2+ ions were selectively adsorbed in the KCuFC crystal 

phase, while other metal ions such as Ru(NO)3+ and ZrO2+ were absorbed in the alginate phase. High uptake percentage 

of 98.6% was obtained by using the dissolved solutions of spent fuel from FBR-JOYO (119 GWd/t, JAEA). <strong>The</strong> alginate 

film enclosing KZnFC was further prepared by using the support of cellulose filter paper, where the Pd2+ ions were 

selectively adsorbed on the KZnFC-MC film. <strong>The</strong> alginate film enclosing insoluble ferrocyanides are predicted for the 

selective separation of Pd2+ as an ion-exchange filter. Thus, the microcapsules enclosing insoluble ferrocyanides are 

effective for the selective separation of Pd2+ from high-level liquid waste (HLLW). 

19) 40113 – An Empirical Model To Determine <strong>The</strong> Modes Of Corrosion Of Carbon Steel 

Toshikatsu Maeda, Masatoshi Watanabe, Seiji Takeda, Shinichi Nakayama, JAEA (Japan) 

Carbon steel is a candidate material for overpack of high-level radioactive waste disposal in Japan. One of its 

expected functions is to avoid groundwater from contacting with vitrified waste form. <strong>The</strong> corrosion rate is a determining 

factor for the overpack lifetime, and it is highly dependent on the mode of corrosion. Carbon steel is an alloy that could 

be attacked by localized corrosion in the form of pitting corrosion or crevice corrosion under certain water chemistries. 

For example, it is known that carbon steel is passivated in solutions above a limiting pH; pH that is referred to as the 

general corrosion/passivation transition pH and is denoted as pHd. Carbon steel corrosion rates depend on the mode of 

corrosion (general or local corrosion), which is determined by the pHd value. Predicting the pHd for carbon steel in near 

field environments is essential to evaluate the expected lifetime of overpacks, especially considering the possibility of 

highly alkaline environments induced by cementitious materials in the disposal facility. In this study, an empirical model 

was developed to determine whether near field environments fall in the passivation or non-passivation domain for carbon 

steel. Using the experimental data obtained by previous studies, the pHd was defined as a function of four factors shown 

in equation (1), where the activity of proton ion ([H+]) for pHd is assumed to be a linear combination of the logarithms of 

the total carbonate concentration (C total), the chloride ion concentration (Cl-), and the limiting current density of 

dissolved oxygen diffusion (iO2), and the inverse of absolute temperature of contacting solution (T). <strong>The</strong> derived 

equation fitted well experimental data from previous studies. 

20) 40124 – Trends in Scenario Development Methodologies and Integration in NUMO's Approach 

Takeshi Ebashi, Katsuhiko Ishiguro,NUMO (Japan); Keiichiro Wakasugi, JAEA (Japan); 

Hideki Kawamura, Obayashi Corporation (Japan); 

Irina Gaus, Stratis Vomvoris, Andrew Martin, Nagra (Switzerland); 

Paul Smith, Safety Assessment Management (UK) 

<strong>The</strong> development of scenarios for quantitative or qualitative analysis is a key element of the assessment of the safety 

of geological disposal systems. As an outcome of an international workshop attended by European and the Japanese 

implementers, a number of features common to current methodologies could be identified, as well as trends in their 

evolution over time. In the late nineties, scenario development was often described as a bottom-up process, whereby 

scenarios were said to be developed in essence from FEP databases. Nowadays, it is recognised that, in practice, the 

approaches actually adopted are better described as top-down or "hybrid", taking as their starting point an integrated 

(top-down) understanding of the system under consideration including uncertainties in initial state, sometimes assisted by 

the development of "storyboards". A bottom-up element remains (hence the term "hybrid") to the extent that FEP 

databases or FEP catalogues (including interactions) are still used, but the focus is generally on completeness checking, 

which occurs parallel to the main assessment process. Recent advances focus on the consistent treatment of uncertainties 

throughout the safety assessment and on the integration of operational safety and long term safety. 

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Abstracts 

21) 40137 – Development of Methodology to Construct a Generic Conceptual Model of River-valley Evolution 

for Performance Assessment of HLW Geological Disposal 

Makoto Kawamura, Kenichi Yasue, Tadafumi Niizato, Shin-ichi Tanikawa, JAEA (Japan) 

In order to assess the long-term safety of a geological disposal system for high-level radioactive waste (HLW), it is 

important to consider the impact of uplift and erosion, which cannot be precluded on a timescale in the order of several 

hundred thousand years for many locations in Japan. Geomorphic evolution, caused by uplift and erosion and coupled to 

climatic and sea-level changes, will impact the geological disposal system due to resulting spatial and temporal changes 

in the disposal environment. Degradation of HLW barrier performance will be particularly significant when the remnant 

repository structures near, and are eventually exposed at, the ground surface. In previous studies, river erosion was 

identified as the key concern in most settings in Japan. Here, therefore, we present a methodology for development of a 

generic conceptual model for performance assessment based on best current understanding of river erosion in Japan. 

Critical considerations that have to be taken into account when interpreting the geological record of past river-valley 

evolutions, as preserved in ancient fluvial deposits, include: 1) the spatial variation in the relative significance of erosion 

and sedimentation at any time between upper- and lower-reaches of rivers originating in mountainous terrain 2) the 

temporal variation in the extent of erosion / sedimentation at any specific location during glacial / interglacial cycles 3) 

the balance between uplift and vertical erosion as a result of the hardness of the riverbed rock 4) the balance between 

vertical and lateral erosion – ranging from formation of narrow gorges to wide meandering flood plains 5) the varying 

duration and intensity (as assessed by sea level change) of past glacial / interglacial cycles. 

Interpretation of the impact of such phenomena at relevant locations in Japan has led to development of a generic 

conceptual model which contains the features typical of mid-reach rivers. This paper presents the methodology to 

develop the conceptual model, identifies the simplifications and uncertainties involved and assesses their consequences in 

the context of repository performance. Details of resultant analyses using this conceptual model will be discussed in 

another paper presented in ICEM’10 by Miyahara et al. 

22) 40162 – Long-term Stability of Bentonite Material Used as Engineered Barrier for Radioactive Waste 

Disposal 

Vera Jedinakova-Krizova, Katerina Videnska, Institute of Chemical Chechnology Prague (Czech Republic); 

Eduard Hanslik, T.G.M. Water Research Institute (Czech Republic) 

Geological concepts for the storage of radioactive waste are based on „multi-barrier“ systems, which are composed 

of several elements: the waste packaging, an engineered barrier incorporating bentonite material and a geological barrier. 

<strong>The</strong> stability of engineered barriers can be affected by long-term pressure and temperature gradients. <strong>The</strong> contribution is 

focused on geotechnical, chemical, geochemical and mineralogical research of bentonite stability with the aim to 

determine the effect of saturation medium composition and loading by heat on bentonite stability. <strong>The</strong> main part the 

research is directed to the experimental results of bentonite and bentonite leachate samples obtained for the bentonite 

interaction under laboratory or in situ experiments. <strong>The</strong> tested bentonite materials represent a series of compact loaded 

bentonite samples taken from the dismantling process of the Mock-Up-Cz experiment, which constitute first vertical 

model of a bench-scale buffer mass test of Czech smectitic clay in the Europe. Mineralogical composition, chemical and 

physicochemical properties including uptake of long-term lived fission products in the form of 99TcO4- and 134Cs+ 

were utilised for the evaluation of the changes caused with the loading of this material. This work aims to characterise the 

basic parameters (ion exchange capacity, mineralogical and chemical composition, specific surface area) of bentonite 

materials, which are important for the retention of the anionic forms of redox-sensitive radionuclides (99Tc, 125I). In 

particular, ion exchange capacity characterises bentonite’s ability to adsorb ions from an aqueous environment. 

Geological concepts for the storage of radioactive waste are based on „multi-barrier“ systems: the waste packaging, an 

engineered barrier incorporating bentonite material and a geological barrier. <strong>The</strong> stability of engineered barriers can be 

affected by long-term pressure and temperature gradients. <strong>The</strong> contribution is focused on geotechnical, chemical, 

geochemical and mineralogical research of bentonite stability with the aim to determine the effect of saturation medium 

composition and loading by heat on bentonite stability. <strong>The</strong> main part the research is directed to the experimental results 

of bentonite and bentonite leachate samples obtained for the bentonite interaction under laboratory or in situ experiments. 

<strong>The</strong> tested bentonite materials represent a series of compact loaded bentonite samples taken from the dismantling process 

of the Mock-Up-Cz experiment, which constitute first vertical model of a bench-scale buffer mass test of Czech smectitic 

clay in the Europe. Mineralogical composition, chemical and physicochemical properties including uptake of long-term 

lived fission products in the form of 99TcO4- and 134Cs+ were utilised for the evaluation of the changes caused with the 

loading of this material. It was found that bentonite material was stable from the point of the long term loading under the 

model conditions applied in the Mock-Up-Cz experiment and only small part of original bentonite material was 

transformed under the occurrence of the newly formed mineral domains (about few weight percent) as result of the long 

term loading bentonite material by pressure and temperature gradients. It can be concluded that listed above newly 

formed mineral phases originated by transformation of bentonite material have no significant influence on the ion 

exchange properties. Migration studies of 99TcO4-and 134Cs+ simultaneously confirmed these results, as well. Similar 

behaviour of bentonite clay barrier can be expected in the long time therefore this material is suitable for the applications 

as engineered barrier in the deep repository of the high level radioactive waste and spent fuel. 

Acknowledgement: This research was supported by the Ministry of Education, Youth and Sports of Czech Republic 

under the project MSM 60461373007 and by Ministry of Industry and Trade under the project FR-TI1/362. 

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Abstracts 

23) 40173 – Development of a methodology for evaluating the performance of host rock for geological disposal 

based on surface-based investigations 

Manabu Inagaki, Keisuke Maekawa, JAEA (Japan); 

Tatsuya Tanaka, Shuji Hashimoto, Obayashi Corporation (Japan) 

In most countries, a step-wise site investigation strategy is applied (or planned) in the site selection process for 

geological disposal of HLW. In Japan, preliminary surface-based investigations such as boreholes and geophysical 

surveys will be followed by detailed investigations, including construction of, and testing in, an underground 

characterization facility. Such investigations assess compliance with legal requirements and provide the understanding of 

geological conditions needed for the development of a repository concept. <strong>The</strong> repository concept, including 

underground design and layout, will depend on the spatial extent of the candidate rock formation(s) and their barrier 

performance under site-specific conditions found deep underground. However, information obtained from initial 

surface-based investigations is inherently limited and includes uncertainties, particularly with regard to the heterogeneity 

of geological characteristics. It is thus important to assess host rock performance in a manner that explicitly considers 

uncertainties in the site characterization. 

In this study, methodology for evaluation of geological conditions has been developed that focuses on determining 

the usable volume of host rock with specific performance characteristics. A test case utilizes the dataset obtained from 

the Horonobe Underground Research Laboratory Project, which is selectively used by taking into consideration step-wise 

data production of a type that would result from preliminary surface-based investigations. Initially, multiple performance 

indices - such as groundwater velocity and solute transport distance -have been defined as key characteristics of the 

geological environment from the viewpoint of barrier performance. Based on the interpretation of the dataset, alternative 

hydrogeological conceptualizations are developed, resulting in several different hydrogeological models. <strong>The</strong> critical 

issue of understanding the impact of uncertainty described in hydrogeological model is addressed by assessment using a 

particle tracking method, which allows quantification of the sensitivity of the performance indices to these uncertainties. 

Finally, a radionuclide migration model was included to further assess the significance of the performance indicators in 

characterizing the impact of hydrogeological uncertainties on the safety of a repository. 

<strong>The</strong> performance indices proposed in this study are seen to be useful to identify requirements and focus 

investigation for the selection of host rocks. Assessing the impact of uncertainties of conceptualization of the 

hydrogeology on the performance indices and nuclide migration then leads to a better understanding of critical issues in 

determining the usable volume of potential host rocks. 

24) 40176 – Structural Integrity Evaluation Approach for PWR Spent Nuclear Fuel 

Yun Seog Nam, Seong Ki Lee, Yong Hwan Kim, Jeon Kyeong Lak, 

Choi Ki Sung, Chang Sok Cho, KNF (Korea Rep.) 

PWR fuel assembly experiences many changes from the time it is manufactured, loaded in the reactor and 

repositioned in the core several times until finally removed from the reactor for the interim storage, reprocessing or final 

disposal etc. Under a severe radiation and a thermo-mechanical condition, this mechanism can alter fuel assembly 

characteristics such as its mechanical properties, geometrical shape, material characteristics etc. Any of these alterations 

which impact spent nuclear fuel (SNF) integrity should be considered to design a cask/canister for its transportation or 

intermediate dry storage. Regarding the cask/canister design, there could be a freedom to design a system that mitigates 

the forces transmitted to SNF and fuel rods. If the storage cask/canister or transport package design prevents or mitigates 

forces transmitted to its contents such that structural integrity is not significantly compromised, the detailed SNF 

properties are necessary to make a decision of the elaborated design parameters, assuming other factors (temperature, 

inert atmosphere, etc.) have been adequately addressed. An approach to those work formations is to analyze mechanical 

characteristics of structural components which are the mechanical properties of grid spring reflecting its irradiation effect, 

SNF strength and SNF structural deformation such as its bow, twist etc. Those informations are also used to evaluate 

hypothetical accident like a drop accident of SNF cask/canister, to select limiting fuel assembly for cask/canister to 

accommodate various kinds of SNFs and to design transportation or storage system for SNFs. Especially, the fuel 

assembly structural properties are a sort of essential data. Thus, in this paper, some approaches to evaluate SNF 

mechanical characteristics are suggested through the existing technical information review, some test data and the 

analysis methodology, and also closely study the mechanical characteristics of a representative SNF for its general 

comprehension. 

25) 40200 – Development of Program Categories to Assess the Radiological Dosage during Spent Fuel 

Transportation 

Suhong Lee , Sangwon Shin, Enesys, Jaemin Lee, Enesys.Co.,Ltd. (Korea Rep.); 

Kiyroul Seong, Jeonghyoun Yoon, KRMC (Korea Rep.) 

Korea is seeking the plan to put spent fuels in interim storage facility as disposal business of low and 

intermediate-Level radioactive waste started. First, <strong>The</strong>re is a strong probability that transport of the spent fuels will be 

arranged in marine transportation. Accordingly, building of transportation risk assessment system in the marine 

transportation comes to the force as a real problem. <strong>The</strong>refore, in this study, I organized the basic categories of program 

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Abstracts 

to develop transportation risk assessment program about specialized transport in marine transportation of spent fuels. <strong>The</strong> 

possibility of radiation exposure in marine transportation consists of the normal condition and the accident condition. <strong>The</strong> 

direct radiation exposure that the workers in loading and crews of the vessels will receive is considered in the normal 

condition. I decide the leakage rate dividing the accident condition that the radioactive materials are spilled into air and 

marine and doing the effect by emission by fire and physical damage. Also, as for the radioactive materials floated in the 

air, calculated applying the atmospheric dispersion factors to them and evaluate the direct radiation exposure and internal 

radiation exposure. As for marine contamination, the radioactive materials move along ocean currents. Generally, there 

can be the direct radiation exposure of ocean activists due to the moving radioactive materials according to the flow of 

the ocean currents in contaminated marine and the direct radiation exposure and the indirect radiation exposure by 

breathing of the subjects near the shores due to the smoke and fog form that occurs by seawater. Also the internal 

radiation exposure by intake contains one of marine life. I set the categories to build the overall programs to build this 

with the comprehensive program that have effects on radiation while the spent fuels are transported combining it with the 

transport conditions or the characteristics of the subjects of radiation exposure and the environmental factors. And 

building of the program will be completed through it. <strong>The</strong> completed program can evaluate various conditions of marine 

transportation after it will be specialized in marine transportation. 

26) 40225 – Exploiting synergies between the UK & Japanese geological disposal programmes 

Ellie Scourse, Atkins (UK); Hideki Kawamura, Obayashi Corporation (Japan); 


<strong>The</strong> early ‘80s UK programme for deep geological disposal of high-level radioactive waste was advanced and at the 

stage of characterising potential sites. When this project was put on hold in the mid ‘80s, much expertise in this field was 

lost. 

In Japan R&D in the ‘80s resulted in major generic safety assessments to demonstrate feasibility in the ‘90s. This 

led to the establishment of NUMO (Nuclear Waste Management Organization of Japan) and the initiation of siting based 

on volunteerism. This novel approach required more flexible methodology and tools for site characterisation, repository 

design and safety assessment. NUMO and supporting R&D organisations in Japan have invested much time and effort 

preparing for volunteers but, unfortunately, no discussions with potential host communities have yet developed to the 

point where technical work is initiated. 

Presently, the UK is moving forward; with the NDA RWMD (Nuclear Decommissioning Agency Radioactive 

Waste Management Directorate) adopting a NUMO-style volunteering approach and a flexible design catalogue. 

Communities have already shown interest in volunteering. <strong>The</strong> situation is thus ideal for collaboration. 

<strong>The</strong> paper will expand on the opportunities for the UK and Japan to benefit from an active collaboration and discuss 

how this can be most efficiently implemented. 

Key Words: UK, Japan, collaboration 

27) 40239 – Realistic Consequence Analysis of River Erosion Scenarios for a HLW epository 

Kaname Miyahara , Makoto Kawamura, Manabu Inagaki, JAEA (Japan); 

Ian G. McKinley, McKinley Consulting (Switzerland); 

Michael J. Apted, Monitor Scientific LLC (USA) 

Uplift and erosion cannot be precluded in most sites in Japan. As no assessment cut-off times have yet been defined, 

erosive radionuclide release scenarios must be developed and analyzed, even if these occur far in the future. Obviously, 

uplift and erosion will cause major disruption of the engineered and natural barriers when the repository nears, and is 

eventually exposed at, the ground surface. In a previous study, a simple linear uplift process was combined with a more 

detailed assessment of river erosion, which was identified as the key erosion process in Japan. This indicated the 

robustness of the reference HLW disposal system: the consequences of erosion of the repository being small when 

compared to the yardstick provided by natural radionuclide fluxes. 

<strong>The</strong> original model was rather simple, but highlighted the importance of cyclic erosion in response to the changing 

environment in glacial and inter-glacial periods. <strong>The</strong>refore, the geological record of such cycles preserved in Japan as 

river terraces has been studied further, with the aim of constructing a more realistic river erosion model. It is evident that, 

although riverbed deepening occurs during glacial cycles, significant sedimentation also occurs and the timing of these 

phases during the cycle differs in the upper and lower reaches of rivers. Unfortunately the situation is particularly 

complicated in mid-reach settings – and these may be most relevant for repositories that avoid mountain locations and 

cannot be established in the highly populated coastal plains. Here interpretation of complex river terrace structures must 

take into account both variations in the length of past glacial/inter-glacial periods and in the resistance of different 

riverbeds to erosion. Further, when the width of river channel is significantly less than that of the valley, meanders 

develop giving sequences of river terraces that are observed only on one side (asymmetric) or on both sides (symmetric) 

of river. Unlike in the upper reaches, in such environments it cannot be assumed that valley profiles are effectively 

constant in time and the gradual progress from eroding mountains towards formation of a peneplain has to be considered. 

This paper describes a conceptual model based on generalization of these observations and resultant consequence 

analyses, again using comparisons with natural radionuclide fluxes. Geological evidence supporting such erosion models 

will be discussed in another paper presented in ICEM’10 by Kawamura et al. 

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28) 40272 – Removal of Fission Products in the Spent Electrolyte Using Iron Phosphate Glass as a Sorbent 

Ippei Amamoto, Masami Nakada, Yoshihiro Okamoto, JAEA (Japan); 

Naoki Mitamura, Tatsuya Tsuzuki, Central Glass Co. Ltd. (Japan); 

Yasushi Takasaki, Atsushi Shibayama, Akita University (Japan); 

Tetsuji Yano, Tokyo Institute of Technology (Japan) 

<strong>The</strong> [3LiCl(59.5mol%)-2KCl(40.5mol%)] eutectic medium used in the pyroreprocessing by the electrorefining method will be 

contaminated by the accumulation of various actinoid elements (An) and fission products (FP) due to prolonged electrolytic operation. 

Though the An can be removed at the pyrocontactor step by extraction using cadmium melt, the FP remain in the eutectic medium, 

which is regarded as a spent electrolyte at certain stage, can cause the rising of the melting point of the electrolytic bath and /or 

lowering the current efficiency. Some measures e.g., its replacement by a virgin medium, etc. should be taken to maintain its stable 

condition. <strong>The</strong> constant replacement of the electrolyte, however, could lead to the generation of enormous volume of high-level 

radioactive waste (HLW). In terms of environmental load reduction and economical improvement, it is desirable to have the spent 

electrolyte purified for recycling by removing its FP. Some technical developments on spent electrolyte treatments have been carried 

out in several countries. One of them is the zeolite sorption of FP following An removal which is being developed in the USA, Japan, 

etc.. Meanwhile, the contaminants precipitation methods by converting contaminants to insoluble compounds are undertaken in Russia 

and a few other countries and are anticipated to reduce the total volume of HLW. In the case of the Russian process, its purpose is to 

remove contaminants such as the minor actinoids (MA) and rare earth elements (REE) from the medium before its disposal. In our 

case, the main objective is to recycle the purified medium, delaying its disposal for as long as possible. It is with this objective in mind 

that this study was undertaken. We have introduced the simple filtration method to remove REE particles which were formed due to 

the conversion of REE chlorides to phosphates. Here, the iron phosphate glass is used as a filtration medium for the removal of FP 

particles. However, some soluble FP such as compounds of alkali-metals, alkaline-earth metals, etc. still remain in the eutectic medium. 

This time around, on an experimental basis, the iron phosphate glass has been used as a sorbent instead, to remove the soluble FP. We 

have obtained some positive results and have intention to incorporate it into the spent electrolyte recycle process as a part of the FP 

separation and immobilization system. 

29) 40295 – Propagation and Interactions of Acoustic Waves in a Waveguide Attached at the Surface of Rock 

Jin-Seop Kim, Kyung-Soo Lee, S. Kwon, KAERI (Korea Rep.); Gye-Chun Cho, KAIST (Korea Rep.) 

<strong>The</strong> dynamic response of a rock mass is inevitable in the systematic long-term monitoring and management in the 

radioactive waste disposal repository. With this point of view, AE (acoustic emission) detection is considered to be a 

promising technique for monitoring the in-situ performance of near-field rock mass. In this study, propagation and 

interactions of guided acoustic waves in a waveguide, which is required to in-situ application of AE monitoring system 

were investigated. <strong>The</strong> changes in acoustic wave amplitude, time delay, frequency variation, and system transfer function 

were measured between the waveguide and a rock sample. Subsequently the waveguide coupling conditions filled with 

epoxy were compared with mechanical type of coupling for the validity of field application. Three type of coupling 

methods were used to compare each other. One is a type of direct contact with granite specimen(Ch.2) which is same 

kind of rock mass with KURT(Korea Underground Research Tunnel) by using a vacuum grease as a reference data. 

Another is a waveguide-aided connection method without any coupling fluid(Ch.1) and the other is a same type with 

Ch.1 except for the use of epoxy resin for a coupling fluid between the gap of a specimen and a waveguide. <strong>The</strong> time 

delay between CH.1 and CH.2 was 0.025 msec and in case of CH.3 and CH.2 0.012 msec. Energy decay ratio of first 

arrival signal for CH.1/CH.2 was 0.68 and 0.81 for CH.3/CH.2. Normalized amplitude attenuation model with a distance 

was successfully obtained. <strong>The</strong> similarity of two signals were quantified using a mathematical tool called mean 

magnitude squared coherence(MSC). <strong>The</strong> mean MSC between CH.1-CH.6, CH.2-CH.6, and CH.3-CH.6 are 0.8727, 

0.9262 and 0.9040 respectively. <strong>The</strong> signal directly attached to the surface of a specimen presents the most similar 

pattern with the applied source signal. While the signals from the epoxy filled waveguide show closer relationship with 

the source than that without filling couplant. Additionally by using of system transfer function, the effect of waveguide is 

apparently shown at the frequency of 118 kHz. <strong>The</strong> results derived from this study can be valuable information for the 

quantitative analysis of signal processing in AE source localization and degree of crack damage in a radioactive waste 

repository. 

SESSION D7: D&D Poster 

1) 40009 – Decommissioning of the KRR-1 & 2 Research Reactors at KAERI; Summary on the Project 

Jin Ho Park, KAERI (Korea Rep.) 

At the Korea Atomic Energy Research Institute (KAERI), two research reactors (KRR-1 and KRR-2) have been 

decommissioned. <strong>The</strong> first research reactor in Korea, KRR-1, was a TRIGA MARK-II type (open pool and fixed core), 

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and its power was 100 kWt at its construction and it was up-graded to 250 kWt by KAERI. <strong>The</strong> second reactor, KRR-2, 

was a TRIGA MARK-III type with an open pool and a movable core and its power was 2 MWt. Its first criticality was 

reached in 1972 and it had been operated for 55,000 hours till the decision to shut it down in 1995. In 1996, it was 

concluded that KRR-1 and KRR-2 would be dismantled. A project was launched for the decommissioning of these 

reactors in January 1997 with the goal of a completion by 2008. <strong>The</strong> total budget for the project is 20.0 million US 

dollars, including the cost for the waste disposal and for the development of the technologies. <strong>The</strong> work scopes during the 

reactor decommissioning project are the dismantling of all the facilities and the removal of all the radioactive materials 

from the reactor site. After confirming the removal of the entire radioactivity by the MARSSIM concept, the site and 

buildings will be released for an unrestricted use after the approval of the regulatory body. This paper summarizes on the 

time tables, technologies, waste generation and treatment, manpower consumption, cost and radiation of work force and 

analyzes the performance on the base of the original plan. 

2) 40075 – Methods of Selected Input Calculation Data Verification And <strong>The</strong>ir Influence On Decommissioning 

Cost In the OMEGA Code 

Frantisek Ondra. Vladimir Daniska, Ivan Rehak, Oto Schultz, DECONTA, a.s. (Slovakia); 

Vladimir Necas, Slovak University of Technology in Bratislava (Slovakia) 

<strong>The</strong> aim of this contribution is development of methodology for verification of selected input calculation data 

(performance unit parameters, work group structure, and duration of time-dependent activities) of the OMEGA Code in 

the individual PSL (Proposed Standardised List) structure parts. <strong>The</strong> OMEGA Code, developed by DECOM, a.s., is used 

for calculation of nuclear power plant decommissioning parameters and decommissioning planning. <strong>The</strong> code represents 

a complex tool modelling a real flow of materials and radioactivity in the whole process of decommissioning, beginning 

from pre-dismantling decontamination and terminating with either final disposal of radioactive waste or release of 

non-contaminated materials to the environment. <strong>The</strong> analytical methodology for evaluation of input data inaccuracy 

impacting on calculation of cost and other output decommissioning parameters was developed. This methodology is 

based on application of coefficients representing calculated cost relative change for contingency adjustment purpose. <strong>The</strong> 

decommissioning process includes significant amount of various technical as well s administrative activities. 

Decommissioning parameters calculation is performed by calculation procedures modelling these decommissioning 

activities. <strong>The</strong> procedures use for calculation a lot of input and unit data. Decommissioning calculation parameters 

projects having been performed using the OMEGA Code showed the necessity of development of a new verification 

module. This module is able to display selected input data used for calculation for either specific inventory database item 

or specific decommissioning activity in the PSL structure. <strong>The</strong> new verification module allows compare selected input 

parameters to reference values including graphic display of differences. <strong>The</strong>re is also a possibility to perform 

recalculation of the some decommissioning option using reference values of input data and consequently to compare 

calculated decommissioning data. <strong>The</strong>se data (e.g. exposure, labor, and cost) and their differences can be compared both 

in tables and graphs, for either specific inventory database item or specific PSL activity. <strong>The</strong> verification module within 

the OMEGA Code brings a new view on impact of selected input data change on calculated decommissioning data. It is 

auxiliary tool for contingency calculation using analytical methods developed last year. 

3) 40078 – Experience of MR and RFT Reactors’ Decommissioning in RRC “Kurchatov Institute” 

Alexander V. Chesnokov,Victor Volkov, Sergey Semenov, Vitaly Pavlenko, Vyacheslav Kolyadin, 

D. Muzrukova, RRC "Kurchatov Institute" (Russia);Artur Arustamov, SUE SIA "Radon" (Russia) 

Research channel-type nuclear reactor «MR» has been developed and commissioned in Russian Research Centre 

«Kurchatov Institute» in 1962-1963. "MR" reactor replaced the old one -"RTF" type reactor. MR is one of 12 units 

located in Kurchatov Institute. It started the operation in 1963, and shut down in 1992. Reactor is located in the water 

pool on the 9 m depth. Nine loop units with different cooling medium are the experimental basis of MR. When MR has 

been shut down; the loop channels were disconnected and placed into water pool. Nuclear fuel has been unloaded and 

moved to the dry repository. Decommissioning design has been developed by Russian Research Centre «Kurchatov 

institute» and JSC «Alliance-Gamma» in 2009 year. According to the decommissioning design the final state of reactor 

MR will come to: - Equipment and systems of reactor shall be dismantled completely; - Technological facilities of 

reactor and MR territory shall be decontaminated and remediated according to the regulations. D&D design has been 

approved by Russian Authority in July 2009 year. In August 2009 the equipment supply and preliminary works have 

been started. Dismantling works are scheduled to start in 2011. Features of the reactor MR which have made an essential 

impact on the development of the project on decommissioning are shown. A status on decommissioning of reactors MR 

and RTF is resulted. <strong>The</strong> work performed on the reactor MR in preparation for decommissioning, which include: removal 

of irradiated fuel from near reactor storage, disposal units of loop channels, acting above the water surface in a pool 

storage are described, identification, sorting and disposal of radioactive objects from gateway of the reactor is carried out. 

<strong>The</strong> technologies for dismantling works were identified and the stages of work on the decommissioning of reactors MR 

and RTF is presented. <strong>The</strong> assessment of radiation effects on the population and the environment during the dismantling 

operations, which is determined by the amount of dismantling operations, the level of radioactive contamination of the 

reactor structures and equipment, as well as the technologies used for their removal is completed. <strong>The</strong> results of the 

estimation of radiation effects on the population and the environment during routine operation to dismantle the 

equipment and in emergency situations are reduced. 

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4) 40126 – Detailed Standardized Decommissioning Parameters Calculation for Larger echnological Aggregates 

and Relevant Buildings in Nuclear Power Plants using the OMEGA Code 

Peter Bezak, Vladimir Daniska, DECONTA, a.s. (Slovakia); Ivan Rehak, DECOM, a.s. (Slovakia) 

Computer code OMEGA, developed by DECOM a.s., is used for evaluation of nuclear facility decommissioning 

activities. <strong>The</strong> Code implements in full extent the standardised cost structure PSL. Decommissioning activities of nuclear 

facility are involved in compact calculation structure. Calculation models a real material and radioactivity flow and 

reflects a radioactivity decay during decommissioning process. Calculation processes material items, which are linked to 

decommissioning procedures (dismantling, demolition and decontamination procedures) in calculation structure. 

Inventory database contains approx 90 standard material items of technological equipment (pipes, valves, tanks etc.), 

approx 50 specific items (pieces, planked components etc.). <strong>The</strong> inventory database also contains 14 building categories 

(masonry, concrete, steel construction etc.). A new task is costing the larger technological aggregates decommissioning 

in nuclear power plants. <strong>The</strong> paper introduces development of larger technological aggregates inventory database. <strong>The</strong>se 

aggregates include: 1. Reactor and its internals 2. Steam generator 3. Pressurizer 4. Refueling machine, etc.. Larger 

technological aggregates decommissioning activities need to be implemented into decommissioning planning and costing 

Code OMEGA. So decommissioning procedures representing these activities have to be developed for technological 

aggregates, and dismantling unit factors need to be set up as well. A proper definition of dismantling techniques and 

workgroups performing the techniques is also important, taking into account presence of activated and contaminated 

materials to be dismantled. Where a dose rate is higher than a limit for manual dismantling, remote dismantling 

techniques are applied. Paper introduces manual and remote techniques available for larger technological aggregates 

dismantling. Dismantling procedures of larger technological aggregates are based on reversed sequence of their 

commissioning. <strong>The</strong> paper also deals with specific dismantling procedures, when equipment is dismantled as a whole and 

moved to a fragmentation facility. <strong>The</strong>re it will be fragmented to smaller parts, put into containers for disposal in 

radioactive waste repository. Decommissioning of larger technological aggregates relates to buildings where aggregates 

are housed in and the paper also deals with their demolition. 

5) 40190 – Dismantling Method of Fuel Cycle Facilities Obtained by Dismantling of the JRTF 

Fumihiko Kanayama, JAEA (Japan) 

<strong>The</strong> Japan Atomic Energy Research Institute Reprocessing Test Facility (JRTF) was the first reprocessing facility 

which was constructed by applying only Japanese technology to establish basic technology on wet reprocessing. JRTF 

had been operated since 1968 to 1969 using spent fuels (uranium metal / aluminum clad, about 600kg as uranium metal 

and 600MWD/T) from the Japan Research Reactor No.3 (JRR-3). Reprocessing testings on PUREX process were 

implemented at 3 runs, so that, 200g of plutonium dioxide were extracted. After JRTF was shut down at 1970, it had been 

used for research and development of reprocessing since 1971. <strong>The</strong> more mature research and development of nuclear are, 

the more opportunity of dismantling of old nuclear facilities would be. JAEA has an experience of full scale of 

dismantling through decommissioning of JPDR. On the other hand, we didn’t have that of fuel cycle facility. Moreover, 

it is considered that dismantling methods of nuclear reactor and fuel cycle facility are different for following reason, 

components contaminated TRU nuclide including Pu, contamination form being many kinds, and components installed 

inside narrow cells. Dismantling methods are important factor to decide manpower and time for dismantling. So, it is 

indispensable to optimize dismantling method in order to minimize manpower and time for dismantling. Considering the 

background mentioned above, the decommissioning project of JRTF was started in 1990. <strong>The</strong> decommissioning project 

of JRTF is carrying out phase by phase. Phase 1; Investigation for dismantling of the JRTF. Phase 2; R&D of 

decommissioning technologies for dismantling of the JRTF. Phase 3; Actual dismantling of the JRTF. <strong>The</strong>re were several 

components used for reprocessing and a system for liquid radwaste storage, and those were installed inside of each of 

several thick concrete cells. <strong>The</strong> inner surfaces of each cell were contaminated by TRU nuclides including Pu. In phase 3, 

components used in reprocessing and a system for liquid radwaste storage were dismantled. Moreover, concrete walls 

(including ceiling) were opened to make entrances in this work. Effective practices for dismantling fuel cycle facilities 

were obtained through these works. On this report, I introduce effective dismantle method obtained by actual dismantling 

activities in JRTF. 

6) 40191 – Computer Simulation of Cryogenic Jet Cutting for Dismantling Highly Activated Facilities 

Sung-Kyun Kim, Kune-Woo Lee, KAERI (Korea Rep.) 

Cryogenic cutting technology is one of the most suitable technologies for dismantling nuclear facilities due to the 

fact that a secondary waste is not generated during the cutting process. In this paper the feasibility of cryogenic cutting 

technology has been investigated by using a computer simulation. In the computer simulation, a hybrid method combined 

with the SPH (smoothed particle hydrodynamics) method and with the FE (finite element) method was used. And also, a 

penetration depth equation, for the design of the cryogenic cutting system, was used and the design variables and 

operation conditions to cut a 10 mm thickness for steel were determined. Finally the main components of the cryogenic 

cutting system were developed on the basis of the obtained design values. 

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7) 40193 – Strippable Core-shell Polymer Emulsion for Decontamination of Radioactive Surface Contamination 

Bum-Kyoung Seo, Bum-Kyoung Seo, Kune-Woo Lee, KAERI (Korea Rep.) 

Strippable coatings are innovative technologies for decontamination that effectively reduce loose contamination. 

<strong>The</strong>se coatings are polymer mixtures, such as water-based organic polymers that are applied to a surface by paintbrush, 

roller or spray applicator. In this study, the core-shell composite polymer for decontamination from the surface 

contamination was synthesized by the method of emulsion polymerization and blends of polymers. <strong>The</strong> strippable 

polymer emulsion is composed of the poly(styrene-ethyl acrylate) [poly(St-EA)] composite polymer, poly(vinyl alcohol) 

(PVA) and polyvinylpyrrolidone (PVP). <strong>The</strong> morphology of the composite emulsion particle was core-shell structure, 

with polystyrene (PS) as the core and poly(ethyl acrylate) (PEA) as the shell. Core-shell polymers of styrene (St)/ethyl 

acrylate (EA) pair were prepared by sequential emulsion polymerization in the presence of sodium dodecyl sulfate (SDS) 

as an emulsifier using ammonium persulfate (APS) as an initiator. Related tests and analysis confirmed the success in 

synthesis of composite polymer. <strong>The</strong> products are characterized by FT-IR spectroscopy, TGA that were used, 

respectively, to show the structure, the thermal stability of the prepared polymer. Two-phase particles with a core-shell 

structure were obtained in experiments where the estimated glass transition temperature and the morphologies of 

emulsion particles. Decontamination factors (DF) of the strippable polymeric emulsion were evaluated with the polymer 

blend contents. <strong>The</strong> decontamination factors obtained for Sr-90 on the disk plate studies were observed the DF values of 

8.9 to 12.8 at all the polymer composition. 

SESSION R4: ER Poster 

1) 40025 – Improvement of quicklime mixing treatment by carbon dioxide ventilation 

Yuki Nakagawa, Hisayoshi Hashimoto, Hitachi Construction Machinery Co., Ltd. (Japan); 

Koichi Suto, Chihiro Inoue, Tohoku University ( Japan) 

This report describes fundamental examination about a quicklime mixing treatment combined with carbon dioxide 

ventilation for the remediation process of soils polluted with volatile organic compounds (VOCs). <strong>The</strong> quicklime mixing 

treatment is widely applied to remove volatile pollutants in soils using 65.17 kJ/mol of the heat from the following 

reaction; CaO+H2O?Ca(OH)2?? To keep higher temperature and to ensure most of VOCs are volatilized, 10 % of 

calcium oxide is usually mixed with soils in this treatment. However, much amount of addition of calcium oxide results 

in higher alkaline soil pH and gives serious damage to the soil ecosystems. To solve this problem, a simultaneous 

ventilation of carbon dioxide during calcium oxide mixing to polluted soil was conducted. <strong>The</strong> formation of calcium 

carbonate according to following reaction produces 92.84 kJ/mol of the heat; Ca(OH)2+H2CO3? CaCO3+2H2O?? It is 

expected that the heat from the above reaction can be used for the treatment and the amount of calcium oxide addition for 

the treatment can be reduced. Laboratory experiments showed that more than half of calcium oxide changed to calcium 

carbonate when carbon dioxide ventilated to the mixed soil sample after 5% of calcium oxide and 5% of water mixed 

with the soil. Maximum soil temperature for this treatment increased same as that for the treatment with 10% calcium 

oxide. Pilot level and operational level experiments confirmed the effectiveness of the simultaneous ventilation of carbon 

dioxide during the quicklime mixing process. 

2) 40166 – Procedure and result of decommission of R&D facility of Uranium fuel 

Hirokazu Tanaka, Masao Shimizu, Ryoji Tanimoto, Kazuhiko Maekawa, 

Shinzo Ueta, Mitsubishi Materials Corporation (Japan); Susumu Tojo, SERNUC Corporation (Japan) 

Mitsubishi Materials Corporation (MMC) had carried out nuclear engineering researches and developments such as 

Uranium fuel development since 1954 in MMC’s research laboratory in Omiya city in Japan. <strong>The</strong> facilities of the 

laboratory had been decommissioned since 1998 to 2005. <strong>The</strong> research activities were transferred to the new research 

laboratories established in different location. At the time of the decommission work, there was no suitable law restriction 

to distinguish the decontaminated waste as radioactive or not. MMC discussed and adopted the pragmatic procedure 

under its own responsibility with consultation of regulatory authority and technical authority outside the company. <strong>The</strong> 

decommissioned material could be divided into two categories. One was research apparatus and building. Another was 

shallow rand soil contacted to the facility. Metal wastes were mainly rose from the apparatus. Metal, concrete, and wood 

wastes were rose from building. At first, these wastes were decontaminated for example by blasting. <strong>The</strong>n the surface 

radioactivity was measured mainly by GM-survey meter and temporally by alpha-survey meter to compare the 

radioactivity criteria. In this case, the radioactivity criteria was decided as the natural radioactivity measured outside the 

facility without any influence of artificial radioactivity. Natural radioactivity varies within a certain range, therefore it 

was measured and treated statistically. Radioactivity of the soils contacted to the facility were measured because there 

was possibility that small amount contaminated liquid had flew out from the facility through the crack of pipes and 

fracture in basement concrete. It was very difficult to decide whether the soil was contaminated or not, because the soil 

naturally contains radioactive nuclides as U-238, Th-232, K-40, etc. Soil from ten areas in the city far away from the 

laboratory were collected and analyzed its radioactivity by the automatic alpha-beta counting system consisted of a 

ZnS(Ag) scintillation detector and a plastic scintillation detector. <strong>The</strong> analyzed data were treated statistically to decide 

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the appropriate value of natural radioactivity. <strong>The</strong> law on decommission work was established later, however there is still 

no suitable quantitative criteria to distinguish uranium contaminated material or not. MMC’s criteria was considered as 

very low, which was close to natural radioactivity. In other words, the facility was decommissioned under very safely 

manner. <strong>The</strong> wastes generated from the decommission work are being held in storehouse build in the same area. 

3) 40210 – Photo-Catalytic Degradation of Industrial Pollutant by Using Nano-technology and Solar energy in 

the Eastern Area of Kingdom of Saudi Arabia 

Osama S.Y. Mohamed, Abdullah A. Al Jaafari, King Faisal University (Saudi Arabia) 

<strong>The</strong> nano-material is one of the most dynamic and fastest growing areas of research in the fields of science, 

engineering, and medicine. Also, in Saudi Arabia, the sun light is available in all year. In the same time, Marine and fresh 

waters can be considered, at present, to consist of a complex mixture of dissolved and non dissolved organic materials as 

well as biological substrates and colloidal aggregates. Following this line, we prepared nano-particles of zinc oxide inside 

Pressurized vessel with temperature controller unit and dried under supercritical conditions in presence of ethanol. <strong>The</strong>se 

nanoparticles have used as photocatalyst in order to destroy colored dye pollutant (Acid green 1) in the presence of sun 

light. In the same trend, we developed zinc oxide nano-particles by doping with aluminum oxide in order to increase the 

efficiency of its visible photo-catalytic activity. <strong>The</strong> experimental results showed that the doped and undoped zinc oxide 

nanoparticles are very active for catalytic degradations of pollutants in presence of sun light. 

4) 40224 – Hydrogen Production from a PV/PEM Electrolyzer System Using a Neural-Network-Based MPPT 

Algorithm 

Abd El-Shafy Nafeh, Electronics Research Institute (Egypt) 

<strong>The</strong> electrolysis of water using a polymer electrolyte membrane (PEM) electrolyzer is a very vital and efficient 

method of producing hydrogen (H2). <strong>The</strong> performance of this method can be significantly improved if a photovoltaic 

(PV) array, with maximum-power-point (MPP) tracker, is utilized as an energy source for the electrolyzer. This paper 

suggests a stand-alone PV/PEM electrolyzer system to produce pure hydrogen. <strong>The</strong> paper also develops the different 

mathematical models for each constituent subsystem. Moreover, the paper develops the suitable maximum-power-point 

tracking algorithm that is based upon utilizing the neural network. This algorithm is utilized together with the action of 

the PI controller to improve the performance of the suggested stand-alone PV/PEM electrolyzer system through 

maximizing the hydrogen production rate for every instant. Finally, the suggested hydrogen production system is 

simulated using the Matlab/Simulink and neural network toolbox. <strong>The</strong> simulation results of the system indicate the 

improved relative performance of the suggested hydrogen production system compared with the traditional case of direct 

connection between the PV array and the PEM electrolyzer. 

5) 40301 – Cement based solidification / stabilization of industrial contaminated soil using various cement 

additives 

Grega E. Voglar, RDA (Slovenia); D. Lestan, Agronomy Department, Biotechnical Faculty, 

University of Ljubljana, (Slovenia) 

A large number of industrial activities produce wastes and contaminants that reach the soil through direct disposal, 

emissions, spills, leaks and other pathways. An increasing number of abandoned industrial sites (brownfields) have 

emerged as a result of weak environmental regulation over decades. Soil clean-up operations (remediations) of 

brownfields, followed by redevelopment is essential to lower the urbanization pressure on arable and other farmland 

(greenfields). <strong>The</strong> town of Celje in central Slovenia has a long tradition of metallurgical and chemical industries, which 

started in 1874 with a zinc smelter and was subsequently expanded to the production of Zn and Pb oxides and Ba salts. In 

1912, the synthesis of H2SO4 and in 1970 the production of TiO2 started. Obsolete manufacturing plants have gradually 

been closed and replaced with new ones in a new location. <strong>The</strong> whole industrial site of “old Cinkarna” was finally 

demolished in 2003, leaving a brownfield area of some 170,000 m2 very close to the city centre and with highly 

contaminated soils, primarily with potentially toxic metals and metaloides (PTMs) and to some extend also with organic 

pollutants. In a laboratory study, 15% (w/w) of ordinary portland cement (OPC), black portland cement (BPC) and 

puculanic cement (PC) combined with various cement additives were used for solidification / stabilization (S/S) of Cd, 

Pb, Zn, Cu, Ni and As contaminated soils from the former industrial site. Soils formed solid monoliths with all cements. 

S/S effectiveness was assessed by measuring the mechanical strength of the monoliths, concentrations of metals in 

deionised water and TCLP (Toxicity Characteristic Leaching Procedure) soil extracts, and mass transfer of metals. 

Concentrations of Cd, Pb, Zn and Ni in water extracts from S/S soils generally decreased, while concentrations of As and 

Cu increased. Concentrations of Cd, Pb, Zn, Cu and Ni in the TCLP extracts from S/S soils were lower than from original 

soils, while the extractability of As from S/S soils increased. Overall, the concentration of metals in deionised water and 

TCLP solution, obtained after extraction of the S/S soils, was below the regulatory limits. S/S greatly reduced the mass 

transfer of Cd (up to 300-times), Pb (up to 53.7-times) and Zn (up to 3843-times). Mass transfer of Ni was generally also 

reduced, while that of Cu and As increased in some S/S soils. Based on the findings of mass-transfer mechanism analysis 

the predominant mechanism of release was surface wash-off of metals otherwise physically encapsulated within the 

cementous soil matrix. 

116

Abstracts 

SESSION M3: EM/PI Poster 

1) 40099 – Removal of Fluorine and Boron from Groundwater Using Radiation-induced Graft Polymerization 

Adsorbent at Mizunami Underground Research Laboratory 

Yosuke Iyatomi, Hiroyuki Hoshina, Noriaki Seko, Noboru Kasai, Yuji Ueki, Masao Tamada, JAEA (Japan) 

High fluorine and boron contents in groundwater are commonly reduced using coagulation and ion-exchange 

treatments. As an alternative, we tested the efficiency of fluorine and boron removal from groundwater using 

radiationinduced graft polymerization adsorbent. <strong>The</strong> durability of the adsorbent was also determined by varying 

groundwater flowthrough rates and repetitive use of the adsorbent. <strong>The</strong> results indicated that it was possible for the 

adsorbent to remove more than 95% of boron and fluorine from the groundwater, and that the performance of the 

adsorbent for boron removal was better than commonly used ion-exchange resin. <strong>The</strong> adsorbent used several times was 

able to remove boron, indicating that the adsorbent can be used for efficient boron removal. 

2) 40184 – <strong>The</strong> Optimized Risk Management of the Waste from NORM and Nuclear Industries - How to 

Harmonize Risk from Various Sources 

Yoko Fujikawa, Kyoto University Research Reactor Institute (Japan); 

Michikuni Shimo, Fujita Health University (Japan); 

Hidenori Yonehara, National Instuitute of Radiological Sciences (Japan); 

Tadashi Tujimoto, Electron Science Institute (Japan) 

We compared the existing regulation on management of radioactive and non-radioactive wastes with the ideal 

legislation procedures for protection of environment. <strong>The</strong> comparison revealed the necessity of risk-based regulation, 

consideration for ethics and cost-effectiveness of the regulation, and optimal usage of regulation resources. In order to 

assess the cost-effectiveness of several different waste disposal options, the concept of disposal cost per unit radiotoxicity 

(mSv or m3) in waste (CPR hereafter) was introduced and calculated. <strong>The</strong> results revealed that current disposal option of 

high level radioactive waste (underground burial) was more cost-effective than that of TENORM and asbestos containing 

waste. 

3) 40205 – Exposure Dose Evaluation of Worker at Radioactive Waste Incineration Facility on KEARI 

Sang Kyu Park, Jong Seon Jeon, Younhwa Kim, Jaemin Lee, NESYS.CO., LTD. (Korea Rep.); 

Ki Won Lee, KAERI, (Korea Rep.) 

In this study, we evaluate the exposure dose of worker by operating at radioactive waste incineration facility that is 

in KAERI(Korea Atomic Energy Research Institute) for safety analysis. <strong>The</strong> incineration facility that is areas are 570 ?, 

annual treatment quantities are 40,000 kg is separated with radiation zone and non-radiation area to installed extra 

equipment for incineration process. <strong>The</strong> incineration facility consist of preparing system, incineration processing system, 

exhaust gas treatment system, ash treatment system and controlled with measurement. Radioactive waste incineration 

facility during normal operation, worker is exposured radiation by the exposure route of external and internal. <strong>The</strong>y are 

exposured by external radiation that was effected by treated waste material and characteristic. In case of internal exposure, 

it is primary fact that is inhalation of contaminated air. <strong>The</strong> evaluation assumption is the next: One is that internal 

exposure is concentration of radioactive material release basis and work is 200 days per year(8 hr per day) because of the 

incineration facility is operated and managed in critical regulation under maximum acceptance air contamination. Result 

of evaluation, maximum exposure dose is 3.07 mSy/y and internal exposure dose is 2.5 mSy/yr according as selecting the 

treated radioacitve waste. <strong>The</strong> exposure dose of assessment result is lower the basis of domestic nuclear law and 

regulation limit in IAEA. 

4) 40209 – Scenario Development for Safety Assessment of Waste Repository for Feasibility Study on 

Transmutation of Spent Nuclear Fuel into LILW using PEACER 

Sung-yeop Kim, Kun Jai Lee, KAIST (Korea Rep.) 

Safety assessment of radioactive waste repository is necessary for feasibility study on transmutation of spent nuclear 

fuel into LILW(Low and Intermediate Level Waste) using transmutation reactor PEACER(Proliferation-resistant, 

Environmental-friendly, Accident-tolerant, Continuable-energy and Economical Reactor). PEACER is a conceptual 

liquid metal fast reactor using Pb-Bi as a coolant. Scenario development is important to the safety assessment for several 

reasons. Scenarios provide the context in which safety assessments are performed. Scenarios influence model 

development and data collection efforts. <strong>The</strong>y have become a very important aspect of confidence building for the 

post-closure safety assessment. In this study, condition of medium depth disposal about 100-200m in granite is 

considered. Waste from PWR and PEACER considering DF(Decontamination Factor) is disposed in this condition. 

Scenario for these concepts are established by screening FEP(Feature, Event, Process) database and benchmarking 

reference safety assessment report. 

117

List of Pre-Registrants 

Kapila FERNANDO Australian Nuclear Science and Technology Organisation Australia 

Lynn TAN Australian Nuclear Science and Technology Organisation Australia 

Robin George HEARD IAEA Austria 

Irena MELE IAEA Austria 

Roman BEYERKNECHT Nuclear Engineering Seibersdorf GmbH Austria 

Wolfgang STUDECKER Nuclear Engineering Seibersdorf GmbH Austria 

Jan DECKERS Belgoprocess Belgium 

Henri VANBRABANT Belgoprocess Belgium 

Paul LUYCX Castor Consulting Belgium 

Lou AREIAS SCK.CEN Belgium 

Alain VAN COTTHEM Technum-Tractebel Engineering Nv Belgium 

Joseph, Ghislain BOUCAU Westinghouse Belgium 

Sheila M. BROOKS Atomic Energy of Canada Limited Canada 

Miklos GARAMSZEGHY Nuclear Waste Management Organization Canada 

Josef PODLAHA Nuclear Research Institute Rez plc Czech Rep. 

Frantisek SVITAK Nuclear Research Institute Rez Czech Rep. 

Karel SVOBODA Nuclear Research Institute Rez, Czech Republic Czech Rep. 

Eduard Josef HANSLÍK T. G. Masaryk Water Research Institute Czech Rep. 

Gérald OUZOUNIAN ANDRA France 

Jean-Guy NOKHAMZON Commissariat à l'énergie atomique France 

Claudio PESCATORE OECD/NEA France 

Florence GASSOT-GUILBERT SGN AREVA GROUP France 

Peter PILLOKAT AREVA NP GmbH Germany 

Christoph Michael STIEPANI AREVA NP GmbH Germany 

Johannes FACHINGER Furnaces Nuclear Applications Grenoble Germany 

Markus Adam HARTUNG NUKEM Technologies GmbH Germany 

Jörg WÖRNER RD-Hanau Germany 

Katharina AYMANNS Research Centre Juelich Germany 

Natalia GIRKE Research Centre Juelich Germany 

Hans-Juergen STEINMETZ Research Centre Juelich Germany 

Yasuo TOMISHIMA AIST Japan 

Motoi KAWANISHI Central Research Institute of Electric Power Industry Japan 

Koji NAGANO Central Research Institute of Electric Power Industry Japan 

Yukihisa TANAKA Central Research Institute of Electric Power Industry Japan 

Masaki TSUKAMOTO Central Research Institute of Electric Power Industry Japan 

Yoshifusa FUKUOKA Chubu Electric Power Co., Inc. Japan 

Masato WATANABE Chubu Electric Power Co., Inc. Japan 

Yoshio KIMURA Chuden CTI Co.,LTD. Japan 

Shinichi HOSOYA DIA Consultants Co., Ltd Japan 

Miyoshi YOSHIMURA DIA Consultants Co., Ltd. Japan 

Chikao Chuck MIYAMOTO EPRI <strong>International</strong> Japan 

Masaaki NAKANO Fuji Electric Holdings Co., Ltd. Japan 

Gen-Ichi KATAGIRI Fuji Electric Systems Co., Ltd. Japan 

Kiyoshi AMEMIYA Hazama Corporation Japan 

Yuki NAKAGAWA Hitachi Construction Machinery Co., Ltd. Japan 

Hirokazu MINATO Hitachi-GE Nuclear Energy Ltd. Japan 

Susumu KAWAKAMI IHI Corporation Japan 

Hirofumi TSUKADA Institute for Environmental Sciences Japan 

Takeshi ISHIKURA Institute of Applied Energy Japan 

Hironobu ABE Japan Atomic Energy Agency Japan 

Ippei AMAMOTO Japan Atomic Energy Agency Japan 

Kenji AMANO Japan Atomic Energy Agency Japan 

Daisuke AOSAI Japan Atomic Energy Agency Japan 

Shiho ASAI Japan Atomic Energy Agency Japan 

Shuji DAIMARU Japan Atomic Energy Agency Japan 

Takahiro HANAMURO Japan Atomic Energy Agency Japan 

Kazumasa HIOKI Japan Atomic Energy Agency Japan 

118

Sohei IKEGAMI Japan Atomic Energy Agency Japan 

Ken-Ichiro ISHIMORI Japan Atomic Energy Agency Japan 

Yuu ISHIMORI Japan Atomic Energy Agency Japan 

Yosuke IYATOMI Japan Atomic Energy Agency Japan 

Gento KAMEI Japan Atomic Energy Agency Japan 

Fumihiko KANAYAMA Japan Atomic Energy Agency Japan 

Akira KITAMURA Japan Atomic Energy Agency Japan 

Takanori KUNIMARU Japan Atomic Energy Agency Japan 

Kensuke KURAHASHI Japan Atomic Energy Agency Japan 

Toshikatsu MAEDA Japan Atomic Energy Agency Japan 

Keisuke MAEKAWA Japan Atomic Energy Agency Japan 

Hitoshi MAKINO Japan Atomic Energy Agency Japan 

Toshiyuki MATSUOKA Japan Atomic Energy Agency Japan 

Yoshihiro MEGURO Japan Atomic Energy Agency Japan 

Kaname MIYAHARA Japan Atomic Energy Agency Japan 

Takashi MIZUNO Japan Atomic Energy Agency Japan 

Shinichi NAKAYAMA Japan Atomic Energy Agency Japan 

Masashi NAKAYAMA Japan Atomic Energy Agency Japan 

Tadafumi NIIZATO Japan Atomic Energy Agency Japan 

Kunio OTA Japan Atomic Energy Agency Japan 

Sidik PERMANA Japan Atomic Energy Agency Japan 

Hiromitsu SAEGUSA Japan Atomic Energy Agency Japan 

Hiroshi SAITO Japan Atomic Energy Agency Japan 

Hiroyuki SANADA Japan Atomic Energy Agency Japan 

Eiji SASAO Japan Atomic Energy Agency Japan 

Haruo SATO Japan Atomic Energy Agency Japan 

Yuji SHIBAHARA Japan Atomic Energy Agency Japan 

Seiji TAKEDA Japan Atomic Energy Agency Japan 

Shinji TAKEUCHI Japan Atomic Energy Agency Japan 

Tadao TANAKA Japan Atomic Energy Agency Japan 

Tetsuya TOKIWA Japan Atomic Energy Agency Japan 

Hiroyuki UMEKI Japan Atomic Energy Agency Japan 

Isao YAMAGISHI Japan Atomic Energy Agency Japan 

Yuhei YAMAMOTO Japan Atomic Energy Agency Japan 

Satoshi YANAGIHARA Japan Atomic Energy Agency Japan 

Hideharu YOKOTA Japan Atomic Energy Agency Japan 

Naoki ZAIMA Japan Atomic Energy Agency Japan 

Tomohisa ZAITSU Japan Atomic Energy Agency Japan 

Tatsujiro SUZUKI Japan Atomic Energy Commission Japan 

Satoshi KARIGOME Japan Atomic Power Company Japan 

Ken-Ichi TANAKA Japan Atomic Power Company Japan 

Kazuhisa YAMAGUCHI Japan Atomic Power Company Japan 

Kazuhiko YAMAMOTO Japan Atomic Power Company Japan 

Keizaburou YOSHINO Japan Atomic Power Company Japan 

Yukihiro IGUCHI Japan Nuclear Energy Safety Organization Japan 

Masami KATO Japan Nuclear Energy Safety Organization Japan 

Yusuke MASUDA Japan Nuclear Energy Safety Organization Japan 

Mamoru KUMAGAI Japan Nuclear Fuel Limited Japan 

Masataka KAMITSUMA Japan NUS Co. Ltd. Japan 

Keiji KUSAMA Japan Radioisotope Association Japan 

Takao IKEDA JGC Corporation Japan 

Atsushi MUKUNOKI JGC Corporation Japan 

Kunihiro NAKAI JGC Corporation Japan 

Kiyoshi OYAMADA JGC Corporation Japan 

Hajime TAKAO JGC Corporation Japan 

Ichizo KOBAYASHI Kajima Corporation Japan 

Toshihiko HIGASHI Kansai Electric Power Company Japan 

Takashi NISHIO Kobe Steel, Ltd Japan 

Yoko FUJIKAWA Kyoto University Japan 

Takumi KUBOTA Kyoto University Japan 

119

Kenji KOTOH Kyushu University Japan 

Takeshi NAKAMURA Kyushu University Japan 

Yuzo YAMASHITA Kyushu University Japan 

Tadahiro KATSUTA Meiji University Japan 

Hirokuni ITO Ministry of Education, Culture, Sports, Science and Technology Japan 

Akira SAKASHITA Mitsubishi Heavy Industries, INC Japan 

Hirokazu TANAKA Mitsubishi Materials Corporation Japan 

Nakazawa TOSHIYUKI Mitsubishi Materials Corporation Japan 

Shinzo UETA Mitsubishi Materials Corporation Japan 

Makoto KAWAMURA Mitsubishi Materials Techno Corporation Japan 

Uchida SHIGEO National Institute of Radiological Sciences Japan 

Keiko TAGAMI National Institute of Radiological Sciences Japan 

Yoji KUSAKA NGK Insulators, Ltd. Japan 

Hitoshi OHATA NGK Insulators, Ltd. Japan 

Katsutoshi TORITA NGK Insulators, Ltd. Japan 

Kojuro YAMAMOTO NGK Insulators, Ltd. Japan 

Yoshihiko HORIKAWA Nuclear Engineering, Ltd. Japan 

Kazuma MIZUKOSHI Nuclear Engineering, Ltd. Japan 

Satoru TAKEDA Nuclear Services Company Japan 

Shigeki AKAMURA Nuclear Waste Management Organization of Japan Japan 

Takeshi EBASHI Nuclear Waste Management Organization of Japan Japan 

Kiyoshi FUJISAKI Nuclear Waste Management Organization of Japan Japan 

Takahiro GOTO Nuclear Waste Management Organization of Japan Japan 

Keisuke ISHIDA Nuclear Waste Management Organization of Japan Japan 

Eiichi ISHII Nuclear Waste Management Organization of Japan Japan 

Kenichi KAKU Nuclear Waste Management Organization of Japan Japan 

Tomio KAWATA Nuclear Waste Management Organization of Japan Japan 

Kazumi KITAYAMA Nuclear Waste Management Organization of Japan Japan 

Satoru SUZUKI Nuclear Waste Management Organization of Japan Japan 

Hiroyuki TSUCHI Nuclear Waste Management Organization of Japan Japan 

Hiroyoshi UEDA Nuclear Waste Management Organization of Japan Japan 

Tatsuya TANAKA Obayashi corporation Japan 

Hidekazu ASANO Radioactive Waste Management Funding and Research Center Japan 

Naoki FUJII Radioactive Waste Management Funding and Research Center Japan 

Takahiro NAKAJIMA Radioactive Waste Management Funding and Research Center Japan 

Ario NAKAMURA Radioactive Waste Management Funding and Research Center Japan 

Hitoshi NAKASHIMA Radioactive Waste Management Funding and Research Center Japan 

Hiromi TANABE Radioactive Waste Management Funding and Research Center Japan 

Takahiro YOSHIDA Radioactive Waste Management Funding and Research Center Japan 

Mito AKIYOSHI Senshu University Japan 

Satohito TOGURI Shimizu corporation Japan 

Yuji IJIRI Taisei Corporation Japan 

Hitoshi MIMURA Tohoku University Japan 

Keisuke SHIRAI Tohoku University Japan 

Yu AOKI Tokyo Electric Power Company Japan 

Satoru KANEKO Tokyo Electric Power Company Japan 

Kazuhiro TAKEI Tokyo Electric Power Company Japan 

Masanori ARITOMI Tokyo Institute of Technology Japan 

Hiroshige KIKURA Tokyo Institute of Technology Japan 

Takatoshi ASADA Toshiba Corporation Japan 

Susumu NAITO Toshiba Corporation Japan 

Yuki YAHIRO Toshiba Corporation Japan 

Koji OKAMATO University of Tokyo Japan 

Satoru TANAKA University of Tokyo Japan 

Jongseon JEON Enesys. Co., Korea Rep. 

Juyoul KIM FNC Technology Co., Ltd. Korea Rep. 

Sukhoon KIM FNC Technology Co., Ltd. Korea Rep. 

Dong-Keun CHO Korea Atomic Energy Research Institute Korea Rep. 

Young-Yong JI Korea Atomic Energy Research Institute Korea Rep. 

Hee Reyoung KIM Korea Atomic Energy Research Institute Korea Rep. 

120

Jin-Seop KIM Korea Atomic Energy Research Institute Korea Rep. 

Bum-Kyoung SEO Korea Atomic Energy Research Institute Korea Rep. 

Gwangmin SUN Korea Atomic Energy Research Institute Korea Rep. 

Yun Seog NAM Korea Nuclear Fuel Korea Rep. 

Sun-Joung LEE Korea Radioactive Waste Management Corporation Korea Rep. 

Jin-Beak PARK Korea Radioactive Waste Management Corporation Korea Rep. 

Ho-Taek YOON Korea Radioactive Waste Management Corporation Korea Rep. 

Rizwan AHMED Kyung Hee University Korea Rep. 

Muzna ASSI Lebanese Atomic Energy Commission(LAEC) Lebanon 

Renate DE VOS Nuclear Research and consultancy Group Netherlands 

Folasade Mulikat LAWAL MULT ENGINEERING Nigeria 

Tamara ZHUNUSSOVA Norwegian Radiation Protection Authority Norway 

Khuram SHAHZAD Peace and Development Organization Pakistan 

Mohamed AMR Qatar University Qatar 

Gheorghe BARARIU R.A.A.N. - S.I.T.O.N. Romania 

Mikhail BOGOD ECOMET-S Russia 

Alexander GELBUTOVSKIY ECOMET-S Russia 

Alexander KOLPAKOV ECOMET-S Russia 

Alexey VOTYAKOV FSUE RosRAO Russia 

Sevastyanov STANISLAV Gazprombank Russia 

Dmitry V. MARININ Institute of Chemistry FEDRAS Russia 

Alexander KOBELEV Scientific and Industrial Association RADON Russia 

Vladimir DANISKA DECOM, a.s. Slovakia 

Peter BEZAK DECONTA, a.s. Slovakia 

Frantisek ONDRA DECONTA, a.s. Slovakia 

Marek VASKO DECONTA, a.s. Slovakia 

Grega E. VOGLAR Regional Development Agency Celje /University of Ljubljana Slovenia 

Helena DANISKOVA Grammar School of St. Michael the Archangel in Piestany Slovakia 

Tomas HRNCIR Slovak University of Technology Slovakia 

Matej ZACHAR Slovak University of Technology Slovakia 

Vladimir NECAS Slovak University of Technology in Bratislava Slovakia 

Michal PANIK Slovak University of Technology in Bratislava Slovakia 

Kamil KRAVARIK VUJE, Inc. Slovakia 

Gerhardus R LIEBENBERG South African Nuclear Energy Corporation (Necsa) South Africa 

Lara DURO Amphos 21 Spain 

Johan ANDERSSON JA Streamflow AB Sweden 

Gunnar HEDIN Westinghouse Sweden 

Ian Gerard MCKINLEY McKinley Consulting Switzerland 

Irina GAUS Nagra Switzerland 

Sven-Peter TEODORI Nagra Switzerland 

Stratis VOMVORIS Nagra Switzerland 

Dorothea SCHUMANN Paul Scherrer Institute Switzerland 

Walter M. HEEP ZWILAG INTERIM STORAGE Switzerland 

Chun-Ping JEN National Chung Cheng University Taiwan 

Fong-In CHOU National Tsing Hua University Taiwan 

Chia Chin LI National Tsing Hua University Taiwan 

Chih Tien LIU Atomic Energy Council Taiwan 

Oleksandr NOVIKOV Chernobyl NPP Ukraine 

Ellie SCOURSE Atkins UK 

Ricahrd Peter SHAW British Geological Survey UK 

Alan SIMPSON Pajarito Scientific Corporation UK 

Henry O'GRADY Parsons Brinckerhoff UK 

Anibal L. TABOAS Argonne National Laboratory USA 

David Walter JAMES DW James Consulting, LLC USA 

Sean Paul BUSHART Electric Power Research Institute USA 

Karen KIM Electric Power Research Institute USA 

Kenzi KARASAKI Lawrence Berkeley National Laboratory USA 

Chin-Fu TSANG Lawrence Berkeley National Laboratory USA 

Frank COCINA Los Alamos National Laboratory USA 

Michael Edward COURNOYER Los Alamos National Laboratory USA 

121

John ZARLING Los Alamos National Laboratory USA 

Yi SU Mississippi State University USA 

Geoffrey John PETER Oregon Institute of Technology - Portland Center USA 

Chase Collins BOVAIRD Pacific Northwest National Laboratory USA 

Danielle JANSIK Pacific Northwest National Laboratory USA 

Mark B TRIPLETT Pacific Northwest National Laboratory USA 

Dawn M. WELLMAN Pacific Northwest National Laboratory USA 

Gary Joseph HANUS Phoenix Solutions Co USA 

Joseph LEGARE S.M. Stoller Corporation USA 

Eric OLSON S.M. Stoller Corporation USA 

Corey Adam MYERS Studsvik, Inc. USA 

Dae Y. CHUNG U.S. Department of Energy USA 

Kurt GERDES U.S. Department of Energy USA 

Gary L. SMITH U.S. Department of Energy USA 

Andrew SZILAGYI U.S. Department of Energy USA 

Catherine HANEY U.S. Nuclear Regulatory Commission USA 

Lawrence Edward KOKAJKO U.S. Nuclear Regulatory Commission USA 

Shawn Rochelle SMITH U.S. Nuclear Regulatory Commission USA 

Joonhong AHN University of California, Berkeley USA 

John Calvin WALTON University of Texas at El Paso USA 

Ronald MORRIS Westinghouse Electric Company USA 

Wallace M. MAYS WM Mining Company LLC USA 

122

ICEM2010 <strong>Conference</strong> Organizers 

<strong>Conference</strong> General Chair 

Satoru Tanaka, University of Tokyo 

Technical Program Chair 

Masanori Aritomi, Tokyo Institute of Technology 

Technical Program Co-Chair 

Anibal L. Taboas, Argonne National Laboratory 

Technical Program Vice-Chair 

Kaname Miyahara, Japan Atomic Energy Agency 

Administrative Chair 

Kazuhiro Takei, Tokyo Electric Power Company 

<strong>Conference</strong> Secretary 

Hiroyoshi Ueda, Nuclear Waste Management 

Organization of Japan 

Exhibition Secretary 

Kiyoshi Fujisaki, Nuclear Waste Management 

Organization of Japan 

ASME Project Directors 

John Bendo, American Society of Mechanical 

Engineers 

Vince Dilworth, American Society of Mechanical 

Engineers 

123 

Track Chairs/Co-Chairs 

Low/Intermediate-Level Radioactive Waste 

Management: 

Track Chair 

Kunihiro Nakai, JGC Corporation 

Track Co-Chair 

Miklos Garamszeghy, Nuclear Waste Management 

Organization 

Spent Fuel, Fissile material, Transuranic and 

High-Level Radioactive Waste Management: 

Track Chair 

Kaname Miyahara, Japan Atomic Energy Agency 


Stratis Vomvoris, Nagra 

Facility Decontamination and Decommissioning: 

Track Chair 

Takeshi Ishikura, Institute of Applied Energy 


Toshihiko Higashi, Kansai Electric Power Company 

Environmental Remediation: 

Track Chair 

Tomohisa Zaitsu, Japan Atomic Energy Agency 


Yuu Ishimori, Japan Atomic Energy Agency 

Environmental Management / Public Involvement / 

Crosscutting Issues: 

Track Chair 

Masaki Tsukamoto, Central Research Institute of 

Electric Power Industry 

Global Partnership and Multi-National Programs: 

Track Chair 

Hiromi Tanabe, Radioactive Waste Management 

Funding and Research Center

The 13th International Conference on Environmental ... - Events

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